2026-10-22T08:26:49Z https://ejournal.brin.go.id/urania/oai

oai:ejournal.brin.go.id:article/2112 2024-12-23T01:35:04Z urania:ART

OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4.8 gU/cm3 DENSITY Sihotang, Juan Carlos Ajiriyanto, Maman Kartaman Nurlaily, Ely Junaedi Br. Ginting, Aslina LEU uranium-silicide post-irradiation examination AlMg2 cladding oxide layer To investigate the performance of AlMg2 cladding in the U3Si2/Al dispersion fuel, oxide layer characterization of AlMg2 cladding of the irradiated U3Si2/Al fuel with 4.8 gU/cm3 density was conducted. The oxide layer on the surface of AlMg2 cladding is one of the changes that occur on the cladding after theU3Si2/Al fuel plate has been irradiated in the RSG-GAS reactor to a burn-up of ∼40%. The characterization and observation of the oxide layer was conducted using SEM (Scanning Electron Microscope) and Energy-dispersive X-ray spectroscopy (EDS). Samples with a size of 3x3 mm were taken from the middle of the fuel plate (middle position). After cutting, metallographic preparation includes mounting, grinding, polishing, and ultrasonic cleaning. SEM preparation was carried out by sputter coating using Au layer. The oxide layer on the AlMg2 cladding has a thickness of 10.3 μm with a uniformly distributed cracks along the oxide layer. Badan Riset dan Inovasi Nasional 2022-10-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2112 10.17146/urania.2022.28.3.6755 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 125-134 2528-0473 0852-4777 ind https://ejournal.brin.go.id/urania/article/view/2112/1306 Copyright (c) 2023 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2298 2024-12-23T01:35:04Z urania:ART

SIMULASI UJI TAK MERUSAK PELAT ELEMEN BAKAR PASCA IRADIASI MENGGUNAKAN METODE PENETRANT TEST Wicaksono, Yusuf Gigih Rahmatullah, Helmi Fauzi Artika, Refa Ismarwanti, Sri Sigit, Rohmad Non-destructive test penetrant test PEB dummy simulation The post-irradiation examination aims to determine the performance of nuclear fuel during irradiation in the reactor. Non-destructive tests at the Radiometallurgical Installation include visual observations, digital X-ray radiography, and ultrasonic tests for inspection of potential defects or anomalies in the fuel plate (PEB). Limitations in the detection of surface defects necessitate other nondestructive test methods, i.e., to simulate the use of the penetrant test method on the PEB dummy. The purpose of this simulation is to determine the ability of the penetrant test method to complement the non-destructive test data that has been previously conducted as well as to review the capability of the hot cell in applying the method. The simulation was carried out using a dummy PEB with artificial defects and used water-washable visible penetrant and solvent-removable visible penetrant. The simulation stages for implementing the penetrant test include precleaning, applying the penetrant, dwell time, removing the penetrant, applying developer, inspection, and post-cleaning. The simulation results with the penetrant test method are able to detect surface defects in the form of cracks and porosity, but cannot detect blister defects. Based on the simulation results, it can be concluded that the penetrant test can be an alternative method to complement the non-destructive test using X-ray and ultrasonic radiographic methods. In addition, to accommodate the application of the penetrant test method in IRM hot cells, it is necessary to develop support facilities while still paying attention to ease of handling and safety systems.  Badan Riset dan Inovasi Nasional 2022-10-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2298 10.17146/urania.2022.28.3.6760 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 143-152 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2298/1425 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2309 2024-12-23T01:35:04Z urania:ART

RADIONUKLIDA CESIUM SEBAGAI INDIKATOR PENENTUAN BURNUP BAHAN BAKAR U3Si2/Al SECARA MERUSAK Br. Ginting, Aslina Yanlinastuti Boybul Supardjo Sungkono Destructive method separation cesium burnup Determination of burn-up using a destructive method was carried out by separation of 137Cs in the irradiated U3Si2/Al fuel element plate (FEP) with a density of 4.8 gU/cm3. Determination of burn-up by the destructive method aims to study the suitability of the burn-up obtained by the non-destructive method or using the Origen Code software. FEP U3Si2/Al with CBBJ code 251 was cut in duplo at the top, middle and bottom with a weight of Top-1= 0.049 gFEP and Top-2=0.058 gFEP, Middle-1= 0.055 gFEP and Middle-2= 0.024 gFEP, and Bottom-1= 0.056 gFEP and Bottom-2=0.075gFEP. FEP samples were dissolved in acid to obtain a fuel solution containing fission products of 134Cs, 137Cs, 235U and other isotopes. The fuel solution was pipetted 1 mL and then transferred from the hot cell to R.135 to separate the 137Cs from 235U by a cation exchange method using zeolite from Lampung. The fuel solution was pipetted 100 μL and put into the vial in duplicate, then 1000 mg Lampung zeolite was added to the cation exchange process. The separation results were 137Cs in solid phase, while uranium and other isotopes in liquid phase. The amount of activity (Bq) of the 137Cs isotope was measured with a Spectrometer- and then used for burn-up calculation. The results of the calculation of the burn-up were 26.9714%; 55,1431%; and 37.8855% for the top, middle and bottom pieces, respectively, with an average burn-up by 40%. These values are not much different from the burn-up obtained by the non-destructive method using gamma scanning, i.e., 24.4%; 52.7% ; and 37.6% for top, middle and bottom positions with an average of 38.23% and a burn-up calculated using the Origen code, which is 40%.  Badan Riset dan Inovasi Nasional 2023-11-08 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2309 10.17146/urania.2022.28.3.6711 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 153-162 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2309/1431 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2502 2024-12-23T01:35:04Z urania:ART

UJI PASCA IRADIASI PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 PADA BURNUP 60%: PENGAMATAN VISUAL, RADIOGRAFI SINAR-X DAN ANALISIS CITRA Sigit, Rohmad Artika, Refa Rahmatullah, Helmi Fauzi Ismarwanti, Sri Br. Ginting, Aslina Supardjo Post-irradiation U3Si2/Al fuel visual inspection x-ray radiography image analysis The development of high density U3Si2/Al fuels has been carried out by increasing the fuel density from 2.96 gU/cm3 to 4.8 gU/cm3. Increasing the density of uranium has an impact on the mechanical integrity of the cladding and its geometric stability, so a post-irradiation examination is required. The post-irradiation examination carried out included visual inspection, X-ray radiographic tests, and image analysis. Visual inspection of the fuel surface was performed using a dedicated periscope in the operating area of hot cell 102 and documented using a camera through the hot cell window. X-ray radiographic test using the voltage and current parameters at 150 kV and 1500 μA, respectively. The radiographic image is further analyzed using ImageJ software. Based on visual inspections, no blisters, swelling, excessive oxide layer, or other surface defects were found. The interpretation of X-ray radiographic images clearly shows a homogeneous fuel profile, no evidence of sub-surface defects, and a burn-up profile that matches the results of the gamma scanning test. Based on non-destructive post-irradiation examination, the overall performance of the PEB U3Si2/Al density of 4.8 gU/cm3 during irradiation on the RSG-GAS core was excellent.  Badan Riset dan Inovasi Nasional 2023-11-08 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2502 10.17146/urania.2022.28.3.6786 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 135-142 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2502/1571 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2503 2024-12-23T01:35:04Z urania:ART

PEMISAHAN DAN ANALISIS 137Cs DALAM LARUTAN ELEMEN BAKAR UJI U3Si2/Al DENSITAS 4,8 gU/cm3 DENGAN BURNUP 60% MENGGUNAKAN METODE PENUKAR KATION Yanlinastuti Nugroho, Arif Br. Ginting, Aslina Noviarty Boybul Haryati, Iis amaludin, Agus Noerpitasari, Erlina Kriswarini, Rosika cesium separation cation exchange U3Si2/Al Lampung zeolite Dowex resin Separation and analysis of cesium isotope in U3Si2/Al fuel with a density of 4.8 gU/cm3 and 60% burnup by cation exchange using Lampung zeolite and Dowex resin has been carried out. The purpose of this research was to determine the weight of 137Cs in the fuel. The U3Si2/Al fuel pieces by the Bottom (B), Middle (M) and Top (T) sections were dissolved in duplicate with 6 M HCl and 6 M HNO3 to obtain a U3Si2/Al fuel solution. The separation of 137Cs started with the measurement 50 μL of fuel solution in 2 mL of 0.1 M HNO3 with a gamma spectrometer to determine the initial weight of 137Cs. The solution was put into a column containing Lampung zeolite and then flowed at a speed of 0.3 mL/minute. The resulting 137Cs isotope was bound to the zeolite and effluent containing other fission isotopes. Dowex resin, Cs carrier, and HCl was added into fuel solution, then put into a column containing R-Cl- resin. The effluent that left the column was put back into the column containing the R-NH4+ resin. The 137Cs isotope that was bound to R-NH4+ resin was eluted using 1 M HCI, then the 137Cs isotope was measured using a gamma spectrometer. The results of the 137Cs isotope measurement shows the weight of each sample with code B1=0.1772 μg/g; B-2=0.1635 μg/g; M-1=0.1395 μg/g; T-1=0.1230 μg/g and T-2=0.1036 μg/g with a recovery of B-1=99.54%; B-2=98.98%; M-1=98.99%; T-1=99.38% and T-2=98.98%. Meanwhile, the average weight of the 137Cs isotope separated by Dowex resin was B-1=0.1575 μg/g; B-2=0.1470 μg/g; M-1=0.1263 μg/g; T-1=0.1140 μg/g and T-2=0.0952 μg/g with a recovery of B-1=94.80%; B-2=95.35%; M1=97.94%; T-1=98.15% and T-2=98.89%. It was concluded that the method for separating 137Cs isotopes in U3Si2/Al fuel solution by using Lampung zeolite and Dowex resin did not have a significant difference. Both methods can be used to separate 137Cs in nuclear fuel. This analysis is used to complement the data to calculate the burnup value.  Badan Riset dan Inovasi Nasional 2023-11-08 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2503 10.17146/urania.2022.28.3.6739 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 163-174 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2503/1572 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2504 2024-12-23T01:36:37Z urania:ART

STUDI PENGARUH PENAMBAHAN YTTRIUM DAN PERLAKUAN PANAS BETA TERHADAP KETAHANAN HIDROGEN PADUAN ZIRCALOY-4-0,1%Mo PADA TEMPERATUR 600°C DAN 800°C Alfatar, Ajra Basuki, Eddy Agus Prajitno, Djoko Hadi Zircaloy-4 hydrogen resistance yttrium variation Electrical energy is a source of energy that is widely used and needed by humans. So that the need for electrical energy will increase along with the increase in the human population. Nuclear Power Plant is one solution to this problem. One of the important components of nuclear reactors, namely nuclear fuel cladding, usually uses a material in the form of a zirconium alloy, but the use of this alloy for a long time will be susceptible to embrittlement due to hydrogen penetration. A series of experiments were conducted to determine the effect of adding yttrium and beta heat treatment on hydrogen resistance and mechanical properties of Zircaloy-4-0.1%Mo-xY alloy (x=0 wt%; 0.5 wt%; 1 wt%). The addition of yttrium causes an increase in the amount of precipitate which is thought to be α-Y while increasing the alloy hardness in both as cast alloys and beta heat treatment. Beta heat treatment decreased the alloy hardness due to an increase in the number of β-Zr phases. Hydrogen resistance can be increased by adding yttrium because it acts as a β-Zr phase stabilizer. Similarly, beta heat treatment can increase hydrogen resistance with the same method, namely increasing the number of β-Zr phases so that the amount of soluble hydrogen increases. The addition of 1 wt% Y to zircaloy-4-0.1%Mo alloy with beta heat treatment showed the best hydrogen resistance with a thickness of 6.24 μm at 600°C and 545.5 μm at 800°C.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2504 10.17146/urania.2022.28.2.6643 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-68 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2504/1573 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2505 2024-12-23T01:36:37Z urania:ART

STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3 Sihotang, Juan Carlos Ajiriyanto, Maman Kartaman Putri, Anditania Sari Dwi Nurlaily, Ely Junaedi Br Ginting, Aslina Supardjo LEU uranium-silicide post-irradiation examination interaction layer fission gas bubbles Uranium-silicide compound fuel dispersed in aluminium matrix (U3Si2-Al) have been used in a large number of research reactors around the world because of its excellent behavior under irradiation. This fuel also provides high uranium density with typical fuel loading up to 4.8 gU/cm3 to compensate for the reduced fissile amount in LEU. To improve the density of current U3Si2-Al (2.96 gU/cm3) used in Indonesian GA Siwabessy Multipurpose Research Reactor, U3Si2-Al dispersion fuel plate with density of 4.8 gU/cm3 (U235 ∼19.75%) had been irradiated in RSG GAS for 175 days at 15 MW power to burnup level of approximately 40%. The characterization was performed using SEM-EDS and optical microscope to study microstructure of the irradiatted fuel, largely the fission gas bubbles and the interaction layer between U3Si2 fuel and Al matrix. The average diameter of the bubbles with diameter from 0.06 to 0.55 μm was 0.21 μm. The interaction layer was identified as U(Al,Si)2,3 with thickness of approximately 1.5 μm. The relatively small fission gas bubbles and the interaction layer didn’t cause swelling on the fuel and the overall performance of the fuel plate was very good.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2505 10.17146/urania.2022.28.2.6670 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 69-78 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2505/1574 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2508 2024-12-23T01:36:37Z urania:ART

KARAKTERISASI MATERIAL INSULASI KABEL LISTRIK TEGANGAN RENDAH DARI PRODUK LOKAL PASCA IRADIASI GAMMA Hutagaol, Antonio Gogo Bayquni, Muhammad Ilham Semi crystalline polymer gamma irradiation degree of crystallinity chemical change The aim of this research is to study the effect of gamma radiation on the semi-crystalline polymer character of local product low-voltage electrical cable insulation material, related to its use in facilities with high gamma radiation. The test samples were irradiated using a Gamma Cell device with radiation doses of 25, 50, 100, 200, 400 and 800 kGy. The degree of crystallinity and compound composition of the test material were analyzed by XRD test and characterized using FTIR. Based on the XRD test, there was no significant difference observed in the diffraction pattern between the non-irradiated test material and the irradiated test material with a large variation in dose. This indicates that there is no major phase change of the crystalline compound contained in the related test sample. The degree of crystallinity of the test sample tends to decrease as the radiation dose increases. The FTIR results show that there is a fluctuating transmittance difference between test samples with different doses of gamma radiation as well as non-gamma irradiated test samples. The presence of increasingly widening spectra at 400 kGy and 800 kGy, at wavenumber 3369 – 3370 cm-1, indicates the presence of O-H stretch clusters. Further tests are needed to strengthen the identification of functional groups, determination of organic compoundsas well as the evaluation of the mechanical character of the test sample, which is also necessary to confirm the suspected disturbance in the polymer-plasticizer association.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2508 10.17146/urania.2022.28.2.6615 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 79-88 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2508/1575 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2509 2024-12-23T01:36:37Z urania:ART

KARAKTERISASI RADIOACTIVE FISSION WASTE (RFW) DARI PRODUKSI RADIOISOTOP LUTESIUM-177 MENGGUNAKAN PROGRAM ORIGEN2 Rosmayani, Lena Rohanda, Anis Abdullatif, R Farzand RFW Radioisotope Lutesium-177 ORIGEN The types and production of radioactive waste continue to increase with the development of the use of technology so that good management is needed so as not to endanger the community or the environment. Production of radioisotopes for nuclear medicine is one of the sources of radioactive waste produced by nuclear reactors. Several radiosotopes produced at the G.A Siwabessy Multipurpose Reactor (RSG–GAS) core include Molybdenum-99, Iodine-125, Iridium-192 and Lutesium-177 (177Lu). The characterization of the 177Lu production waste was carried out using the ORIGEN2 computer program with the aim of obtaining the characteristics of the RFW waste to obtain an appropriate waste treatment technique. In this study, a characterization of radioactive waste was carried out on the production of the radioisotope 177Lu from Target Lu2O3 at various CIP positions at a power of 15 MWt and 30 MWt with a long irradiation time at each power level for 8 and 12 days. In radioisotope production, the target is irradiated in the RSG-GAS core irradiation facility, the target is placed in a quartz ampoule which is then placed in a capsule in aluminum. Then the aluminum inner capsule is inserted into the outer capsule and placed into the irradiation position. After irradiation targeting and then taking 177Lu radioisotope. Sampling of radioisotopes after irradiation forms several types of waste. One of them is radioactive fission waste (RFW) as a by-product with different properties. Apart from the target, radioactive waste can also be generated from the target capsule. Target irradiation makes the target capsule made of quartz and aluminum also activated and becomes radioactive. Therefore, it is necessary to calculate the activity concentration of the waste produced by the radioisotope 177Lu. The results of the characterization of RFW waste using the ORIGEN2 computer program with flux variations at various CIP positions at a power of 15 MWt and 30 MWt with an irradiation period of 8 and 12 days obtained a total concentration of waste activity from the production of 177Lu radioisotope which has an activity concentration between 1,06x1016 – 1,24x1016 Bq/g. Therefore, the RFW radioactive waste produced by 177Lu is classified as moderate radioactive waste based on the Government Regulation on radioactive waste management and requires careful management to ensure safety.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2509 10.17146/urania.2022.28.2.6672 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 89-100 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2509/1577 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2511 2024-12-23T01:36:37Z urania:ART

ESTIMASI PANAS GAMMA PADA PRODUKSI RADIOISOTOP Lu-177, Ir-192, DAN Au-198 DI TERAS MOLIBDENUM RSG-GAS Riskiana, Dewi Nur Rohanda, Anis Abdullatif, R Farzand Gamma heating UMo RSG-GAS radioisotopes Gamma heating is an important issue related to the irradiation safety of a reactor and its irradiated samples. Gamma heat is generated from the interaction of gamma energy with the target material. The gamma energy produced from the reactor has different gamma energy determined by the fuel type. Uranium molybdenum (UMo) is a future fuel that has several advantages over uranium oxide (UO2) and silicide (USi), one of which to increase the reactor operatingtime. For this reason, a computational study of the RSG-GAS gamma heat with UMo fuel was carried out in several radioisotope productions such as the Lu-177, Ir-192, and Au-198. This study uses the ORIGEN code to calculate the gamma energy produced for UMo fuel types. Gamma heat estimation uses the Gamset program, designed and tested to calculate gamma heat in RSG-GAS. UMo-fueled gamma heat calculation results in RSG-GAS are 4.85 W/g ~ 8.69 W/g. This result is smaller than the gamma heat of uranium silicide (USi), which is around 9.27 W/g ~ 13.3 W/g. The radioisotope Lu-177 has the highest gamma heat of 8.69 W/g, followed by Au-198 and Ir-192, which are 5.89 W/g and 7.12 W/g, respectively. The gamma heat of the three radioisotopes produced by RSG-GAS with UMo fuel does not exceed the maximum gamma heat (20 W/g) that has been determined to ensure that it is safe for the reactor and samples.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2511 10.17146/urania.2022.28.2.6654 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 101-112 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2511/1578 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2512 2024-12-23T01:36:37Z urania:ART

PEMODELAN DINAMIK PENDINGINAN BAHAN BAKAR NUKLIR BEKAS REAKTOR RISET SECARA NATURAL KONVEKSI PADA PROTOTIPE DRY CASK STORAGE Istavara, Arifin Ratiko Pratama, Hendra Adhi Nasruddin Spent nuclear fuel natural convection canister dry cask storage This study specifically aims to test the feasibility of dry cask storage design as a solution for storing spent nuclear fuel (SNF) in Indonesia. Due to the limited storage space in wet type storage, this research designed, experimented and simulated simultaneously. The canister design test varies the voltage from 50 V to 125 V with closed conditions to determine the response of the canister to the voltage which is proportional to the decay heat of SNF. Experiments using dry cask storage ventilation varying the voltage from 100 V to 175 V which aims to test dry cask storage against convection natural canister cooling. Theoretical calculations and simulations using software were also carried out as a comparison of the experimental results in terms of naturally convection cooling and thermal resistance. The experimental results show the response of the canister design to function properly, the greater the voltage supplied, the higher the canister temperature, 50 V to 125 V responds from 33.4°C to 56.6°C. The canister cooling test by natural convection showed good results, the dry cask ventilation opening and closing method showed a decrease in canister temperature at a voltage of 100 Volts of 16.1 °C and 14.8 °C at 125 V. The canister temperature comparison between experiments, theoretical calculations and simulations at 175 V are 44.9 °C, 49.7 °C and 65 °C respectively, as for velocity are 0.20 m/s, 0.25 m/s and 0.39 m/s respectively. The results of theoretical calculations and simulations are slightly higher than experiments, this is due to heat loss to the environment during the experiment. The simulation results obtained temperature contours and natural convection flow behavior in the air gap, indicating the dry cask storage design is functioning properly.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2512 10.17146/urania.2022.28.2.6639 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 113-124 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2512/1579 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2537 2024-12-23T01:36:55Z urania:ART

EFFECT OF Al, Zr AND Mo ON CORROSION RESISTANCE OF AlXCrFeNiMo AND AlXCrFeNiZr (X=1, 1.2 AND 1.4) AS FUEL CLADDING MATERIALS Firmansyah, Teguh Dilasari, Bonita Setiawan, Jan Hadi Prajitno, Djoko Fuel cladding high entropy alloy corrosion The high entropy alloy of AlxCrFeNiM (with x = 1, 1,2 and 1,4; M = Mo and Zr) was successfully synthesized using powder metallurgy technique with sintering process at 1000 oC in an inert atmosphere. These alloys were designated as fuel cladding for research reactor with high U-density fuel such as U-Mo. One of the critical in-service properties of nuclear fuel cladding is its corrosion behavior. In this study, some properties of the HEA of AlxCrFeNiM such as the phases, microstructures, hardness, and corrosion resistance in 3 wt.% NaCl solution at room temperature were investigated. The results show that the phases in the HEA of AlxCrFeNiMo are FeNi, AlNi and Mo. The HEA of AlxCrFeNiZr has more complex phases compared to the AlxCrFeNiMo. The microstructure of HEA samples show fine grains with some micropores that imperfect the solidification during the sintering process. The hardness value of the HEA of AlxCrFeNiMo has a trend of decreasing as the x value increases. The opposite trend occurs to the HEA of AlxCrFeNiZr that the hardness value increases with increasing x value. The lowest hardness value is Al1.4CrFeNiMo at 262.2 HV, and the highest hardness value of Al1.4CrFeNiZr is at 756.7 HV. The corrosion rate of the AlxCrFeNiMo does not show a specific trend with increasing x value; however, the AlxCrFeNiZr shows decreasing value with increasing x value. The lowest value for the all-HEA samples is 0.20 mmpy for the Al1.4CrFeNiZr. The results of hardness and corrosion tests show that the Zr element combined with the Al element affects not only its hardness but also its corrosion resistance. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2537 10.17146/urania.2022.28.1.6624 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2537/1600 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/2543 2024-12-23T01:36:55Z urania:ART

EFEK NITROCARBURIZING PADA MATERIAL LOW CARBON STEEL TERHADAP KETAHANAN KOROSI DAN KEKERASAN Al Afghani, Fajar Hermawan, Erwan Shaleha, Siti Nitrocarburizing corrosion hardness low carbon steel The use of steel is currently experiencing a fairly rapid growth. Low carbon steel is widely used in radiometallurgical installation, especially for structural components of conveyors, chains, gears and so on. Its use is increasingly massive so that various kinds of developments in steel materials have been carried out. One of them is to improve the quality of steel, so it can be used for a long time. One method that can be used to improve the quality of steel is nitrocarburizing process. This research focused on the effect of the nitrocarburizing process on corrosion rate. The nitrocarburizing process was carried out on several samples with different temperatures but the same gas. Increasing the temperature in the nitrocarburizing process resulted in harder specimen surface. The increase in hardness was obtained based on the temperature difference which increased from 192 HV for the sample before dinitrocarburizing to 793 HV for the 450 ºC nitrocaburizing sample. In addition, corrosion resistance increases with increasing nitrocarburizing temperature. The nitrocarburizing process causes the diffusion of C and N atoms to form a compound layer, diffusion layer, and transition layer consisting of a layer of ε-Fe2-3N, γ’- Fe4N, Fe2(N, C)1-z, making the surface layer of the material become hard and resistant to corrosion.  Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/2543 10.17146/urania.2022.28.1.6596 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-20 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/2543/1601 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/6611 2024-12-23T01:32:51Z urania:ART

ANALISA DAN INTERPRETASI DATA HASIL UJI TAK MERUSAK DALAM RANGKA PENGEMBANGAN BASIS DATA POTENSI CACAT PADA PELET UO2 Artika, Refa Fauzi Rahmatullah, Helmi Gigih Wicaksono, Yusuf Ismarwanti, Sri Sigit, Rohmad Non-destructive testing UO2 pellets visual testing digital X-ray radiography database The development of nuclear fuel for both research and power reactors has been the focus of Research Center for Nuclear Fuel Cycle and Radioactive Waste Technology (PRTDBBNLR) research from fabrication, pre-irradiation, irradiation, and post-irradiation examination phases. As the only post-irradiation examination facility in Southeast Asia, Radiometallurgy Installation (IRM) is required to be ready to evaluate the performance of nuclear fuel, one of them is the short pin PWR-fuel containing UO2 pellets which is currently in the irradiation process. One of the efforts made by IRM is to develop non-destructive testing techniques, including methods, analysis, interpretation, and inventory of potential defects that may occur systematically in short pin PWR-fuel containing UO2 pellets. The aim of this research is to inventory methods and data of potential defects in UO2 pellets for a post-irradiation examination database. Characterization was performed on five samples of pre-irradiated UO2 pellets with variations in sintered/raw, good/defect conditions, and differences in uranium enrichment (3%, 4%, and 5%) using visual test methods and digital X-ray radiography. Visual examination can provide interpretation results regarding the conditions on the surface of the UO2 pellets. Digital X-ray radiography was operated at a voltage of 150 kV and a current of 1800 µA to produce a digital image then be carried out an image enhancement process using an image processor to interpret the UO2 pellets. Digital X-ray radiography method and image enhancement process adds information of UO2 pellets data. Cracks and surface holes of UO2 pellets can be mapped. The difference in uranium enrichment can be observed from the gray value graphic. Information from this interpretation will be useful as an initial database of interpretation methods to determine the condition of the UO2 pellets in the irradiated short pin PWR-fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/6611 10.17146/urania.2023.29.2.7001 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 63-70 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/6611/4941 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0/

oai:ejournal.brin.go.id:article/6615 2024-12-23T01:32:51Z urania:ART

KOMPARASI PERFORMA MONITOR RADIASI GAMMA DALAM PEMANTAUAN RADIASI REAL–TIME Arifianto, Rokhmat Sulaiman, Robi Slamet Parulian Hutabarat, Ricard Cahya Ciputra, Roni Dian Indrastomo, Frederikus Radiation detector linear regression ANOVA The National Research and Innovation Agency has a research facility used for research on nuclear materials that have potential radiation hazards that can endanger workers. To reduce the potential radiation hazard, radiation monitoring is carried out routinely by workers. To optimize radiation monitoring, three (3) radiation monitors were developed using 1 (one) scintillation (S) radiation detector and 2 (two) GM detectors (GM and GMT) which can later be installed and monitor radiation in real-time. Testing of the 3 radiation monitors developed was carried out in this study to determine the measurement accuracy of each radiation monitor. The tests carried out were linear regression tests to obtain a conversion model from units of counts per second to units of microsivert per hour. In addition, ANOVA testing was conducted to see if there was a significant difference between the dose rate measurements from the three radiation monitors compared to the calibrated surveymeter. Tukey HSD testing was carried out to test each of the developed radiation monitors and compared with a calibrated surveymeter. The results of linear regression testing between the surveymeter (GS) and the three radiation monitors obtained a coefficient of determination above 0.95. ANOVA testing conducted found that there was a significant difference in the average radiation dose rate measurement results from each radiation monitor. Tukey HSD test results show that only one of the radiation monitors has an average radiation dose rate measurement value that is not significantly different from the average radiation dose rate measurement value of the calibrated surveymeter. Therefore, the GMT radiation monitors that have been developed are radiation monitors that are suitable for use in radiation monitoring at nuclear materials research facilities. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/6615 10.17146/urania.2023.29.2.6981 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 71-86 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/6615/4945 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0/

oai:ejournal.brin.go.id:article/7777 2024-12-23T01:32:51Z urania:ART

VERIFIKASI PENGUKURAN RADIOAKTIVITAS 153Sm MENGGUNAKAN DOSE CALIBRATOR MULTITIPE Pertiwi, Ligwina Dita Maskur, Maskur Yanto, Yanto Fairusya, Nuha Ekaningsih, Nuri Jannati Wahyu Dose calibrator Verification Radioactivity 153Sm Measurement verification is crucial to ensure measurement results validity. In this study, we have done verification on 153 Sm radioactivity using five different types of dose calibrators dose calibrator Atomlab 100, Atomlab 300, Atomlab 400, Atomlab 500, and Capintec CRC-tR5. The parameters of verification process including accuracy, precision, linerity, Limit of Detection (LOD), and Limit of Quantification (LOQ). The dose calibrator Atomlab 300 was established as the standard instrument to examine the others type of dose calibrator accuracy. Verification results showed all the instruments have an error of less than 3%. However, the t-test results exhibited only the dose calibrator Atomlab 400 that had tvalue less than the corresponding ttable value. In contrast, Atomlab 100, Atomlab 500, and Capintec CRC tR5 produced a tvalue greater than a ttable value. In the precision test, all types of dose calibrator gave a satisfactory precision results. All five-dose calibrators demonstrated good linearity for measuring radioactivity within the range of 20-140 mCi. The LOQ values for each dose calibrator were as follows dose calibrator Atomlab 100 = 7,88 µCi, Atomlab 300 = 12,65 µCi, Atomlab 400 = 17,32 µCi, Atomlab 500 = 15,06 µCi, and Capintec CRC tR5 = 18,6 µCi. As well as the LOQ values of the dose calibrator were determined as calibrator Atomlab 100 = 2,74 µCi, Atomlab 300 = 1,14 µCi, Atomlab 400 = 8,2 µCi, Atomlab 500 = 6,79 µCi, and Capintec CRC tR5 = 7,38 µCi. These showed that five dose calibrator types offer reliable measurement validity with a 95% confidence level. Nevertheless, it is necessary to employ a correction factor for Atomlab 100, Atomlab 500, and Capintec CRC tR5. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7777 10.17146/urania.2023.29.2.6838 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 87-94 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7777/5936 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7778 2024-12-23T01:32:51Z urania:ART

ADSORPSI URANIUM TERLARUT OLEH KOMPOSIT AMMONIUM DIHYDROGEN PHOSPHATE/ZEOLIT Purwanto, Yuli Saputra, Dwi Luhur Ibnu Mirawaty, Mirawaty Sugiharto, Untung Wati, Wati Rachmadetin, Jaka Artiani, Pungky Ayu Composite Zeolite Ammonium dihydrogen phosphate Dissolved uranium Adsorption Zeolite is one of natural aluminosilicate minerals that widely studied for adsorbent for liquid waste treatment. This mineral has also been studied for treatment of waste generated from nuclear industries, including adsorption of uranium from liquid waste. The treatment of this kind of waste is critical in various steps of nuclear fuel cycle. Modification of zeolite, which can be carried out physically or chemically, was reported can increase the adsorption performance of the zeolite. In this research, the capability of zeolite was increased by adding Ammonium Dihydrogen Phosphate (ADP) to form an ADP/zeolite composite for uranium adsorption. The objective of this study is to determine the adsorption performance of ADP/zeolit composite for aqueous uranium. Natural zeolite from Lampung was used as starting material. The composite material is prepared by mixing Ammonium Dihydrogen Phosphate and zeolite in a boiling flask. The resulting composite was characterize using XRD and used for adsorption of aqueous uranium by batch method. The results show that the predominant mineral in the natural zeolite was clinoptilolite. Modification of the zeolite by ADP was succeed as was shown in the XRD pattern. The experimental results are shown by the kinetics of uranium adsorption by the ADP/zeolite composite which takes place quickly and reaches equilibrium in less than 16 minutes. The adsorption kinetics of uranium onto the ADP/zeolit composite was found to follow pseudo-second order. The optimum pH for the adsorption was 7. The uranium adsorption capacity of ADP/zeolit composite was higher than that of natural zeolite. The highest uranium adsorption was obtained for the composite with composition ratio of ADP and zeolite was 1:1. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7778 10.17146/urania.2023.29.2.6974 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 95-104 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7778/5937 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7780 2024-12-23T01:32:51Z urania:ART

ADSORPSI URANIUM MENGGUNAKAN Na DAN Zr – MONTMORILLONITE Saputra, Dwi Luhur Ibnu Purwanto, Yuli Setiawan, Risdiyana Hikmat, Moh. Cecep Cepi Fadilla, Nayandra Dias Ayuningtyas, Pratiwi Musthofa, Malik Mirawaty, Mirawaty Rachmadetin, Jaka Adsorption Na-montmorillonite Zr-montmorillonite Zirconium (IV) oxychloride octahydrate uranium solution The montmorillonite can be used as an alternative for adsorption treatment of uranium waste. Cations in the montmorillonite interlayer can be used as ion exchange in uranium. Uranium liquid waste can be generated from uranium extraction process for molybdenum-99 radioisotope extraction and also in the yellow cake dissolution process.  The purpose of this study is to determine the adsorption of uranium using Na montmorillonite and Zr montmorillonite.  Modification of Na-montmorillonite was carried out using zirconium (IV) oxychloride octahydrate (ZrOCl2.8H2O). The Na and Zr-montmorillonite were characterized using XRD and XRF analysis. Parameters of uranium adsorption were determined by batch experiments with contact time variations of 10, 30, 60, 120, 180, 300, 360, 420 and 480 minutes and pH variations of 3, 5, 7 and 9. Kinetics of uranium adsorption were studied using pseudo first-order and pseudo second-order adsorption. The results show that the optimum pH Na-montmorillonite was pH 5 of 93,68% and Zr-montmorillonite was pH 7 of 96,52 %. The ratio of Kd values in Na and Zr-montmorillonite were 23.42 x 103 and 48.26 x 103 mL/g, respectively. The results of the first-order and second-order pseudo-plot curves show that Na and Zr-montmorillonite were in the second-order plot and constant (k2) were 0.000354 and 0.000458 (g/mg.min), respectively. In conclusion, modification of montmorillonite using Zr increased the uranium adsorption and kinetics. Therefore, this Zr-montmorillonite can be used as an alternative adsorbent for treatment of uranium liquid waste. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7780 10.17146/urania.2023.29.2.6979 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 105-114 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7780/5938 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7781 2024-12-23T01:32:51Z urania:ART

ANALISIS PERILAKU ISOTERM, TERMODINAMIKA, DAN KINETIKA ADSORPSI STRONTIUM MENGGUNAKAN TiO2 – ZEOLIT LAMPUNG Sholahudin, M. Fiqri Purwaningsih, Henny Noerpitasari, Erlina Boybul, Boybul Yanlinastuti, Yanlinastuti Isotherm Adsorption Strontium TiO2 Zeolite After being used in the reactor, nuclear fuel produces several fission products, one of which is strontium (90Sr). Strontium isotopes can be used as raw material for making batteries, but the strontium is still mixed with fission products or other isotope so a separation process is needed. One method of separating strontium is by adsorption. In this research, the Lampung TiO2-zeolite mixture was evaluated for its performance as an adsorbent so that it can be used for strontium adsorption. Lampung Zeolite was initially activated using concentrated sulfuric acid to clean pores that still contained impurities. Then the zeolite is mixed with TiO2 using the stir casting method. The Lampung TiO2-zeolite adsorbent obtained was then characterized using Fourier Transform Infrared (FTIR), X-Ray Diffraction Spectrometer (XRD), and Scanning Electron Microscope (SEM). After that, adsorption tests were carried out using isotherm, thermodynamic and kinetic models. The formation of TiO2-zeolite which was characterized using FTIR spectroscopy was shown by the presence of a Ti-O-Si absorption peak in the Lampung zeolite after TiO2 was added. The results of analysis using XRD show that the Lampung zeolite is clinoptilolite and mordenite, and the TiO2 used has an anatase phase. The results of analysis using SEM indicate that TiO2 is distributed on the surface of the Lampung zeolite. The adsorption isotherm model for the TiO2-zeolite adsorbent fulfills the Langmuir isotherm model. Thermodynamic studies show that the adsorption process is endothermic and not spontaneous, and the kinetics of strontium adsorption follows pseudo second order. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7781 10.17146/urania.2023.29.2.6840 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 115-124 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7781/5939 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7793 2024-12-23T01:38:34Z urania:ART

PENGARUH ROL DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6 Octadamailah, Saga Nurjaya, Dwi Marta Supardjo Suryaman, Ganisa Kurniati Cold rolling Mechanical properties Al 6061-T6 Tensile strength Elastic modulus PENGARUH PROSES PENGEROLAN DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6. Proses pembuatan pelat tutup dan pelat bingkai untuk pelat elemen bakar reaktor G.A. Siwabessy dilakukan dengan cara dirol dingin, sehingga dicapai ketebalan yang diinginkan. Pada penelitian sebelumnya, proses pengerolan dingin pada material AlMg2 menyebabkan perubahan sifat mekanik, dimana material menjadi lebih ulet dan kekuatan tariknya menurun. Pada bahan bakar densitas tinggi, material kelongsong Al 6061-T6 akan digunakan untuk menggantikan material kelongsong sebelumnya, yaitu AlMg2. Tujuan dari penelitian ini adalah untuk mengetahui perubahan sifat mekanik dari material Al 6061-T6 hasil pengerolan. Untuk melihat perubahan sifat mekanik yang terjadi, dilakukan uji tarik, pengamatan mikrostruktur dan struktur kristal. Hasil uji tarik menunjukkan bahwa kekuatan tarik, modulus elastis dan kekuatan luluh cenderung naik seiring dengan besarnya derajat reduksi ketebalan. Meskipun demikian, kekuatan luluh dari hasil pengujian cukup rendah. Rasio antara kekuatan luluh dan kekuatan tariknya hanya 0,7. Dari hasil pemindaian menggunakan difraksi sinar-X diketahui bahwa proses pengerolan dapat menyebabkan penghalusan ukuran kristalit. Kesimpulan dari penelitian ini secara umum adalah proses pengerolan dapat meningkatkan sifat mekanik material Al 6061-T6. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7793 10.17146/urania.2021.27.1.6339 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7793/5941 https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7796 2024-12-23T01:34:08Z urania:ART

PENGARUH KONDISI PENYIMPANAN TERHADAP KETAHANAN KOROSI DRUM BAJA KARBON WADAH LIMBAH RADIOAKTIF Saputra, Dwi Luhur Ibnu Aisyah, Aisyah Setiawan, Risdiyana Artiani, Pungky Ayu Heriyanto, Kuat Rachmadetin, Jaka Radioactive waste Carbon steel Welded materials Corrosion The Radioactive Waste Management Installation (IPLR) in the Serpong Nuclear Area is Indonesia’s only radioactive waste management facility. This facility has treated various types of waste using carbon steel containers, including low and medium-contamination solid waste. Radioactive waste containers must have good corrosion resistance so that the waste container can survive a long storage period. Research on the corrosion of carbon steel radioactive waste containers has been carried out. This study aims to obtain data on the corrosion rate of welded materials radioactive waste container and their parent metals under disposal conditions. In the disposal facility, there is a potential for water seeps in so that the interaction with the waste container may occurred. This condition will increase the corrosion rate. In this study measured the corrosion rate in various corrosion media, namely SP4 water, demineralization water, cemented solution, and bentonite solution. The results showed that the corrosion rate of the welded container was higher than the corrosion rate of its parent metal in various corrosion media. The highest corrosion rate is SP4 water media, which is 7.86 and 6.78 times compared to demineralization water. Meanwhile, the corrosion rate in demineralization water media, bentonite water, and cement water is not significantly different. The corrosion rate is an essential parameter of radioactive waste containers because waste containers with high corrosion resistance will be able to confine radionuclides for a long period time. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7796 10.17146/urania.2023.29.1.6770 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7796/5943 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7797 2024-12-23T01:34:08Z urania:ART

SAFETY INTEGRITY LEVEL ASSESSMENT AT URANIUM EVAPORATOR AND DEPOSITION VESSEL IN NON NUCLEAR REACTOR INSTALLATIONS Oktavianto, Putra Kundari, Anis Noor Saputra, Ade Abdurrosyid, Imam Sholikhah, Munisatun Saputra, Andri Non-Reactor Nuclear Installations PFD SIF SIL SIS The operation of non-reactor nuclear installations that use nuclear material in the process must be ensured safely during the process. One of the assessments of the safety level of the protection system that has been owned by the installation is using the Safety Integrity Level (SIL) which assesses the safety level of the protection system based on the value of the risk reduction factor that the protection system can achieve. The uranium evaporator and deposition vessel at the Experimental Fuel Element Installation (EFEI) is one of the installations that uses nuclear material in the process so it is necessary to assess the SIL of these 2 vessels. The piping and instrumentation diagram (P&ID) is used to determine the SIL value and Safety Instrumented System (SIS) component are installed in evaporator and deposition vessel. Maintenance data and OREDA (Offshore Reliability Data) are used to determine the failure rate. After knowing and determining the installed SIS components, then determining the architecture vote of the Safety Instrumented Function (SIF) based on the P&ID diagram, so that it can be known that the installed SIS uses vote 1oo1, 1oo2, 1oo3, or the appropriate vote. The installed SIF architecture vote will determine the equation used to calculate the Probability Failure on Demand (PFD). The total PFD obtained is adjusted to the SIL table to find out what SIL level the installed protection system is at. The value of the safety level of the protection system with SIL assessment in the evaporator vessel obtained 2 protection systems with SIL values of level 2 all and in the deposition vessel obtained 2 protection systems with SIL values of level 1 and level 2. The SIL value in the evaporator and deposition vessel analyzed has not reached level 3 or 4, so it is necessary to add SIF to the SIS protection system to increase the SIL value until the SIL value is obtained between level 3 or 4 because safety in the operation of non-reactor nuclear installations is absolute. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7797 10.17146/urania.2023.29.1.6986 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-20 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7797/5944 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7798 2024-12-23T01:34:08Z urania:ART

ANALISIS KRITIKALITAS PROSES HYDRIDING SERTA PENYIMPANAN PADUAN UMo DAN UZr SEBAGAI KANDIDAT BAHAN BAKAR REAKTOR RISET MENGGUNAKAN OPENMC Octadamailah, Saga Isnaini, Anissa Sigit, Rohmad U7Mo U6Zr criticality hydriding monte carlo openMC Current research on research reactor fuel is focused at increasing uranium density to produce higher neutron flux in order to increase radioisotope production as well as efficiency and accuracy in various material tests. Research and developments of nuclear fuel in Indonesia in order to increase uranium density is the development of UMo and UZr. One of the processes for making UMo and UZr powder is through a hydriding-dehydriding process which involves enriched uranium, and high pressure hydrogen gas. This combination presents the potential for criticality accidents that must be anticipated. Criticality calculations were carried out using OpenMC software, which is a nuclear physics simulation that uses the Monte Carlo method. In this simulation, two types of alloys are used, namely U7Mo and U6Zr. Simulations were carried out using 10,000 particles and 44 batches as parameter settings. From the simulation results, the keff value for the U7Mo and U6Zr alloy are 0.0297 and 0.0314, respectively. Both values are far below the critical value, indicating that the potential for criticality accidents is very low. In the storage simulation of hydriding products, there is the potential for criticality accidents if the storage bottles are placed close together without any distance. To anticipate this risk, each bottle must be given a safe distance and there needs to be a limit on the permitted mass of uranium per unit volume. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7798 10.17146/urania.2023.29.1.6976 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 21-32 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7798/5945 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7799 2024-12-23T01:34:08Z urania:ART

POTENSI LINGKUNGAN KAWASAN NUKLIR SERPONG DAN EVALUASI KESELAMATAN UNTUK TAPAK FASILITAS BOREHOLE DISPOSAL LIMBAH SUMBER RADIASI BEKAS Setiawan, Risdiyana Megasari, Kartini Sucipta, Sucipta Setiawan, Andry Potensial Environment Site Safety Borehole disposal Disposal of used radiation sources (DSRS) still faces several obstacles, among others, because DSRS generally have a long half-life, high levels of gamma-photon radiation, difficult options for returning to the country of manufacture, and activity that exceeds the limit for shallow land disposal. For this reason, it is necessary to develop a disposal system for DSRS waste with small-scale facilities that are economical, meet worker, community and environmental safety standards. The answer to this problem is the concept of borehole disposal (BHD). This study focuses on evaluating the environmental potential of the Serpong Nuclear Area (KNS) for the BHD facility site and its safety according to the IAEA standard, namely borehole disposal of disused sealed sources (BOSS). Environmental evaluation of KNS is carried out by characterization and evaluation of geological and non-geological environmental safety for BHD sites according to IAEA standards. In general, the results of the evaluation of environmental parameters against the criteria for the disposal site indicate a suitability. However, there are several parameters that need to be given a technological solution (engineered barrier) so that the site's function as a natural barrier can be fulfilled. Some of these technological solutions include waste packaging (capsules and containers), buffer and backfill materials, borehole linings (casing), basement plug and upper/cover plug installation. The results of the safety assessment using Borehole Disposal Concept (BDS) scoping tools show that the BHD facility concept at the KNS site is included in the green zone which means it is safe, with a maximum total dose value of 7.2e-9 Sv/y in 46,678 y. Using the BOSS standard system with the addition of an engineered barrier, the site location at KNS meets the criteria and has the potential for placement of the DSRS waste borehole disposal facility. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7799 10.17146/urania.2023.29.1.6717 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 33-44 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7799/5946 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7800 2024-12-23T01:34:08Z urania:ART

INVESTIGASI MODIFIKASI LUBANG PENGISIAN GAS HELIUM PADA DESAIN END PLUG BATANG UJI BAHAN BAKAR NUKLIR MENGGUNAKAN ANALISIS COMPUTATIONAL FLUID DYNAMICS Ghufron, Hanif Bahar, Shofwan Rianto, Sugeng Vankabo, Praditya Rahmadi, Gagad Hutapea, Odi Buana Triarjo, Triarjo Setiawan, Iwan CFD Simulation PRTF Nuclear fuel A modified design of the helium gas filling hole in the end plug of the PRTF test rod was investigated using CFD simulation method. The simulation is conducted to predict whether helium gas can flow properly into the cladding through the modified design. From this simulation, it is expected that data will be obtained on the effect of changes in the shape and dimensions of the helium gas filling hole. The data will then be used as a guide to determine whether the modified design can function better so that it can replace the original design. The input data required for CFD simulation are the geometry of the model, the type of material used, and the condition parameters for the preparation of the welding process such as pressure and time. Then, the CFD simulation process is carried out in accordance with the conditions that occur in the TIG orbital welding machine. The simulation results show that the modified design is able to flow helium gas into the cladding twice as fast as the original design. In addition, the simulation results also show that the modified design is as stable as the original design in maintaining helium gas pressure at 1.5 bar.   Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7800 10.17146/urania.2023.29.1.6956 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 45-52 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7800/5947 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7802 2024-12-23T01:34:08Z urania:ART

EFEKTIVITAS DESORPSI CESIUM OLEH SURFAKTAN KATIONIK BERDASARKAN STRUKTUR HIDROFILIKNYA PADA MONTMORILLONIT K10 Wijayanto, Herry Nakashima, Satoru Cs desorption Cationic sufactant Montmorillonite interlayer This study discusses the efficiency of cesium (Cs) desorption from montmorillonite K10 samples (MMTK10) previously contaminated with Cs (MMTK10/Cs) using cationic surfactant desorbents, benzyldodecyldimethylammonium bromide (BDAB) and dodecyltrimethylammonium bromide (DTAB), as well as HCl as a desorption control. The aim of this research is to explore the effects of the different hydrophilic structures of cationic surfactants on Cs desorption from MMTK10, where BDAB has a larger hydrophilic structure compared to DTAB. The results show that cationic surfactants, particularly at high concentrations of 20-50 mM, exhibit higher desorption efficiency compared to HCl, with efficiencies reaching 80-90%. This mechanism is associated with the ability of cationic surfactants to form micelles when their concentration exceeds the critical micelle concentration (CMC) or at high concentrations, which facilitates ion exchange with Cs on the surface of MMTK10/Cs. There is no significant difference in Cs desorption efficiency between BDAB and DTAB, even though the hydrophilic structure of BDAB is larger. This contrasts with previous studies on other montmorillonites, where a larger hydrophilic surfactant structure led to greater opening of the montmorillonite interlayer due to the entry of the larger surfactant, resulting in greater Cs desorption. The lack of a difference in Cs desorption efficiency between the two surfactants in this study is due to the specific characteristics of MMTK10 clay itself, which does not have the ability to expand or enlarge the interlayer due to heat treatment during its production before being sold in the market. Additionally, the similarity in the length of the hydrophobic tail chains between BDAB and DTAB contributes to the identical Cs desorption results from both surfactants. In conclusion, while the nature and structure of cationic surfactants influence the Cs desorption process, the specific characteristics of the used montmorillonite or clay also play a crucial role in this desorption mechanism. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7802 10.17146/urania.2023.29.1.6962 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 53-62 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7802/5948 Copyright (c) 2023 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7803 2024-12-23T01:36:55Z urania:ART

IMAGE ENHANCEMENT CITRA DIGITAL SHORT PIN PWR FUEL BERISI PELET UO2 ALAM PASCA UJI RADIOGRAFI SINAR-X MENGGUNAKAN PROGRAM IMAGEJ Artika, Refa Sigit, Rohmad Ajiriyanto, Maman Kartaman Yulianto, Tri Image enhancement Digital X-ray radiography Short pin PWR ImageJ Interpretation Non-destructive test using digital X-ray radiography is one of the important series in performance evaluation of nuclear materials and fuels. Image enhancement or image quality improvement is needed for further analysis of digital X-ray radiography images which tend to have lower resolution. Image enhancement process on digital x-ray radiographic images needs to be carried out to provide images that are easier to interpret so that the resulting data can be used as a basis for evaluating performance after irradiation. The tests were performed at a distance of 40 cm against the detector at voltage and current of 120 kV and 1000 μA. The image enhancement was later performed using ImageJ, which involves the process of aligning the image contrast, edge enhancement, processing histogram value, and adjusting image colors. The resulted data were costumized according to the information to be obtained from the digital X-ray radiography test such as dimensional data, boundaries between pellet, and histogram ilustrating material or thickness differences. Further, this information can be used as data input in mapping the concept of enhancement of image of short pin PWR fuel contains UO2 pellets. Output image produced from the enhancement process not only has a better resolution and quality to facilitate the process of analysis and interpretation but also can be used as a standard method to perform image enhancement of digital X-ray radiography images of short pin PWR fuel contains UO2 pellets after irradiation. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7803 10.17146/urania.2022.28.1.6617 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 21-28 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7803/5949 Copyright (c) 2022 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7804 2024-12-23T01:36:55Z urania:ART

PENGARUH PENAMBAHAN LARUTAN H3BO3 DAN LiOH TERHADAP PERILAKU KOROSI MATERIAL KELONGSONG ZIRCALOY-2 DALAM MEDIA AIR BEBAS MINERAL Al Afghani, Fajar Hermawan, Erwan Putri, Anditania Sari Dwi Nurlaily, Ely Zircaloy-2 Corrosion rate H3BO3 LiOH Cladding The primary coolant of the PHWR type reactor is controlled chemically by the addition of boric acid and LiOH to anticipate corrosion of the cladding material of Zr-2. This study aimed to analyze the effect of H3BO3 and LiOH solutions on the type and rate of of zircaloy-2 (Zr-2) corrosion in demineralized water. The corrosion test was carried out at room temperature to eliminate the influence of high temperature and pressure on the electrochemical process. In this study, the corrosion rate of Zr-2 was observed in a demineralized water solution with varying concentrations of H3BO3 and LiOH. The observation was done using Tafel method, where the addition of H3BO3 at concentrations of 10 ppm, 100 ppm, 500 ppm,1000 ppm and 2000 ppm resulted in corrosion rate of 17.29 x 10-3 mpy, 18. 51 x 10-3 mpy,20.82 x 10-3 mpy, 22.71 x 10-3 mpy and 23.29 x 10-3 mpy, respectively. The addition of 2000 ppm of H3BO3 and LiOH with varied concentrations of 1 ppm, 2 ppm, 3 ppm, 4 ppm results in decreasing of corrosion rate sequentially, namely 22.71 x 10-3 mpy, 21.88 x 10-3 mpy, 21.41 x 10-3 mpy, 21.39 x 10-3 mpy, whereas the addition of 5 ppm LiOH causes the corrosion rate to increase to 21.45 x 10-3 mpy. It can be concluded that the addition of LiOH and H3BO3 affects the corrosion rate of zircaloy-2 in demineralized water. Higher concentration of H3BO3 causes the corrosion rate to increase meanwhile, the addition of LiOH causes the corrosion rate to decrease to a concentration of 4 ppm. However, the addition of LiOH at a concentration of 5 ppm causes the corrosion rate to increase. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7804 10.17146/urania.2022.28.1.6609 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 29-36 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7804/5950 Copyright (c) 2022 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7805 2024-12-23T01:36:55Z urania:ART

NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE Rajagukguk, Santo Paulus Bakhri, Syaiful Surbakti, Tukiran Wigner-Seitz approximation WIMS-D5 code MOX fuel Doppler reactivity The WIMSD-5B thermal reactor lattice cell code is used in many laboratories for research reactor calculations and power reactors. The program uses the Wigner-Seitz approximation for cell pin calculations. The approximation has been widely applied to the pin of UO2 cells and has shown good results in previous studies but can produce incorrect results when used for pin cells in MOX fuels. This paper investigates the use of the WIMS-5B code to calculate the neutron multiplication factor and depletion for MOX fuel pin cells. Calculations were performed using the WIMSD-5B code updated with the ENDF-BVIII.0 library. The outer scattering boundary condition was used to overcome the effect of the Wigner-Seitz approach on the lack of MOX fuel. Results of this study indicates that most of the results obtained using ENDF-BVIII.0 are better than ENDF-BVII.1, and this is in line with expectations. The difference in the maximum k-inf value obtained from this library occurs in the fuel that has the greatest enrichment. On the other hand, the addition of the outer scattering limit improves the results obtained using ENDF-BVIII.0, causing a slight improvement for other libraries. This shows that by using appropriate libraries and the addition of the scattering outer limit, the Wigner Seitz approximation for the MOX pin cell pins in WIMS-D5 can yield quite accurate results. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7805 10.17146/urania.2022.28.1.6610 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 37-44 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7805/5951 Copyright (c) 2022 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7806 2024-12-23T01:36:55Z urania:ART

EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR Suwoto Dwijayanto, R. Andika Putra Luthfi, Wahid Zuhair Fuel density Reactivity coefficients Kinetic parameters MCNP6 ENDF/B-VII.1 Few decades ago, a large number of nuclear reactors were designed to use HEU as the fuel. But the use of HEU is being discouraged since it can be used as a nuclear explosive material which makes its proliferation potential. Most of the HEU-fueled nuclear reactors in the world are either closed or converted into other types that have LEU fuel with 235U enrichment below 20%. To extend lifetime, LEU fuel with high density is developed. The change in fuel density from low to high will also change core neutronics and thermal hydraulics of the reactor, and as a result, it affects the transient response of the reactor. This paper studies the effects of fuel density on reactivity coefficients and kinetic parameters of pebble bed reactor through several calculations with MCNP6 code combine with ENDF/B-VII.1 continuous energy cross-section nuclear data library. The overall calculation results show that the Doppler temperature coefficient (DTC) increases with increasing fuel density, but the moderator temperature coefficient (MTC) decreases due to hardening of neutron spectrum. Kinetic parameters such as effective delayed neutron fraction (βeff), prompt neutron lifetime(ℓ) and neutron generation time (Ʌ) which significantly reduced with increasing fuel density will strongly affect the reactor control and safety. The results of this work conclude that the selection of 9–15 g/cm3 fuel density should be considered carefully given that its effect on reactor controllability. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7806 10.17146/urania.2022.28.1.6625 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 45-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7806/5952 Copyright (c) 2022 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7807 2024-12-23T01:40:53Z urania:ART

PERCOBAAN PEMBUATAN PELAT ELEMEN BAKAR U-7Mo-xSi/Al DALAM UKURAN MINI Supardjo Kadarjono, Agoeng Isfandi Susanto, Yatno Dwi Agus Permana, Setia Guswardani U-7Mo-xSi fuel alloy high uranium density research reactor The purpose of this study is to obtain the optimized process parameters in the fabrication of a U-7Mo-xSi/Al mini plate complying with the specifications. The process flow started with synthesizing U-7Mo-xSi alloys (x = 1%, 2%, and 3%) using depleted uranium and continued to preparing U-7Mo-xSi powder, U-7Mo-xSi/Al fuel core, and U-7Mo-xSi/Al fuel element plates. The quality of the product was determined by analyses/tests including tests of raw materials and products from each end of the process by the appropriate type of test which is commonly applied to the manufacturing process of plate type fuel. The quality of U-7Mo-xSi alloy ingots, U-7Mo-xSi powder, and U-7Mo-xSi/Al fuel core was satisfactorily good and in compliance with the specifications, indicating that the applied process parameters were appropriate. The obtained fuel element plates, however, have not satisfied the requirements especially the thickness of the cladding, where there are still several measurement points whose thickness is <0.25 mm. To obtain cladding thickness according to the required specifications, it is necessary to experiment with other rolling parameters or replacing the cladding material with material of higher hardness. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7807 10.17146/urania.2019.25.3.5697 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 141-152 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7807/5953 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7808 2024-12-23T01:40:53Z urania:ART

PENGARUH TOOL ROTATION SPEED TERHADAP SIFAT MEKANIK SAMBUNGAN ALUMINIUM PADUAN 6061-T6 PADA PROSES FRICTION STIR WELDING Tarmizi Indrawan, Riki Irfan Friction Stir Welding (FSW) Fusion welding Rotational speed Tool Aluminium alloy 6061-T6 Friction stir welding is a newly developed welding process in 1991, and various studies have been continuously conducted to determine parameters that can produce optimum welded joints as an alternative to fusion welding process which still shows some imperfections. The purpose of this research is to determine the effect of tool rotation speed, which is one of the important parameters in friction stir welding of 6061-T6 aluminum alloy plate with 6 mm thickness,  on mechanical properties of the welded joints, where variation of rotation speed applied was 910 rpm, 1175 rpm, 1555 rpm, 1700 rpm and 2000 rpm. The results show that the welded joint resulted from a speed of 910 rpm, 1175 rpm and 1555 rpm indicate no defects and satisfy the acceptance criteria of AWS D17.3 standard.  The most optimum mechanical characteristic was found for welded joint with rotation tool speed of 910 rpm because grain coarsening and dissolution of precipitate were not too significant when compared to other rotational speeds. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7808 10.17146/urania.2019.25.3.5703 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 153-164 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7808/5954 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7809 2024-12-23T01:40:53Z urania:ART

KARAKTERISASI MIKROSTRUKTUR, KEKERASAN, KOMPOSISI KIMIA DAN STRUKTUR KRISTAL LAPISAN PADA PERMUKAAN BAJA Sungkono Ajiriyanto, Maman Kartaman Ismarwanti, Sri Sigit, Rohmad Surface layer Steel Microstructure Hardness Chemical composition Crystal structure Steel and alloys are used as structural material for nuclear reactor components and supporting equipments of hot cells. During the operation of nuclear reactors or hot cell facilities, structural materials suffer from tensile, compressive, and frictional stresses in high radiation environments and thus reduce the life time of the components. One solution for this issue is surface coating of steel with materials having high hardness, corrosion and wear resistance. The aim of this research is to obtain the characteristics of microstructure, thickness, hardness, chemical composition and crystal structure of layers on steel surface. The method used was observation with optical microscope for microstructure, vickers microhardness tester for hardness, spark spectrometry for chemical composition and X-ray diffractometry for crystal structure. The results shows that the base metal microstructure consists of fine equiaxial grains with ferrite and pearlite phases, where the metal-layer interface is clearly visible, and the compact and homogeneous layer has an average thickness of 624.071 µm. The hardness of the layer is 943 VHN. From the chemical composition it is identified that the base metal is S45C carbon steel, and the surface layer is a stable compound of chromium oxide (Cr2O3) with hexagonal crystal structure. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7809 10.17146/urania.2019.25.3.5708 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 165-172 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7809/5955 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7810 2024-12-23T01:40:53Z urania:ART

PENGARUH DIAMETER NOZZLE TERHADAP DIAMETER GEL PADA PROSES GELASI EKSTERNAL KERNEL CERIA STABILIZED ZIRCONIA (CSZ) Nurwidyaningrum, Kristanti Sediawan, Wahyudi Budi Perdana, Indra External gelation Diameter Nozzle In the production of coated particles that meet the requirements for high temperature gas reactor fuels, one of quality control requirements for the kernel is diameter control. The study was conducted by varying the diameter of the nozzle for predicting the size of gel diameter obtained from external gelation process by correlating variables of feed density, feed viscosity, feed solution tension and nozzle diameter quantitatively with dimensional analysis method in the form of dimensionless numbers. Fabrication of CSZ kernel was done by preparing the feed solution, which consists of a mixture of ZrO(NO3)2.2H2O and Ce(NO3)3.6H2O, urea, tetrahydrofurfuryl alcohol (THFA) and poly vinyl alcohol (PVA). External gelation is performed by dripping sol feed solution through a nozzle into gelation column containing ammonium hydroxide solution to form CSZ gel. Droplet formation occurs when the feed solution passes by the nozzle. The diameter of the gel formed can be calculated by correlating nozzle diameter and physical properties of the feed solution using dimensional analysis. Gel diameter of the experimental results was measured using an optical microscope. The diameter of the gel obtained from the dimensionless number equation shows results that are not too different from the diameter of the experimental gel with some deviations in the data. Data of diameter measurement of experimental gel and data of calculation by dimensionless equation show that diameter of the nozzle is directly proportional to diameter of the obtained gel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion https://ejournal.brin.go.id/urania/article/view/7810 10.17146/urania.2019.25.3.5592 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 173-182 2528-0473 0852-4777 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7811 2024-12-23T01:40:53Z urania:ART

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR Pinem, Surian Surbakti, Tukiran Kuntoro, Iman Neutronic Thermal-hydraulic parameter Control rod effect TRIGA 2000 Silicide fuel Analysis of neutronic and thermal-hydraulics parameters of whole operation cycle is very important for the safety of reactor operation. During the reactor operation cycle, the position of the control rods will change due to reactivity changes. The purpose of this study is to determine the effect of control rods position on neutronic and thermal-hydraulics parameters in relation to the safety of reactor operation of the TRIGA 2000 reactor using silicide fuel of MTR plate type. Those parameters are power peaking factor, reactivity coefficients, and steady-state thermohydraulic parameters. Neutronic calculations are performed using a combination of WIMSD/5 and Batan-3DIFF codes and for thermal-hydraulics the calculations are done using WIMSD/5 and MTRDYN codes. The calculation results show that the reactivity coefficient values are negative for all control rod positions both at CZP and HFP conditions. The MTC value decreases when the control rod is inserted into the active core while the FTC value increases. The total ppf results and temperature in steady-state rise when the control rods are inserted of into the active core whereby the maximum value occurs at the position of the control rods of 20 cm from the bottom of the active core. The calculation results of ppf, reactivity coefficient, and thermal-hydraulics parameters lay below safety limits, indicating that the TRIGA 2000 reactor can safely use U3Si2-Al silicide fuel as a substitute fuel for cylindrical type fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7811 10.17146/urania.2019.25.3.5576 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 183-192 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7811/5956 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7812 2024-12-23T01:40:53Z urania:ART

PEMODELAN DAN SIMULASI KINERJA PIN UJI BAHAN BAKAR PWR DENGAN UO2 DIPERKAYA Mutiara, Etty Yulianto, Tri Dewayatna, Winter PRTF Simulation PWR pin FEMAXI Fuel performance Modeling and simulation of irradiation of PWR type test fuel of mini pin in RSG GAS reactor has been done using thermomechanics computer code of FEMAXI-6. The simulation was conducted by varying fuel enrichment under irradiation to predict fuel performance. The expected result from the simulation is the availability of simulation data on fuel behavior under simulated irradiation with different enrichment, neutron fluxes along the fuel, length of time and burn-up level at the beginning of contact between fuel pellets and cladding (Pellet-Cladding Mechanical Interaction,PCMI). The simulation data will provide valuable information for real irradiation experiment planning in RSG GAS reactor. The input for the simulation includes geometry and materials and operation condition such as neutron fluxes and linear heat rate. The calculations include heat generated by the fuel, heat transferred out of the fuel, axial and radial  distribution of  temperatures along the fuel, fission gas release and also pressure generated in fuel pin for whole length of the fuel pin. Temperature distribution calculation from thermal analysis module is applied to detailed mechanical analysis module to calculate gap width between the pellet and the cladding. The simulation shows a trend for reduction of contact time between fuel and cladding with increasing enrichment level at a same reactor power level because higher enrichment will cause higher linear heat rate. While PCMI does not occur on natural uranium fuel even for 100% burnup at full power level, fuel pin containing 5% enriched uranium causes the occurance PCMI at 900 hours of irradiation despite a reduce of power level by half with only 4% burnup. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7812 10.17146/urania.2019.25.3.5692 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 193-204 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7812/5957 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7813 2026-04-29T04:24:32Z urania:ART

PELAPISAN PERMUKAAN BAJA TAHAN KARAT AISI 304 DENGAN KHROM OKSIDA MENGGUNAKAN METODA SPUTTERING Sungkono Setiawan, Jan Isfandi Aziz, Ikhwanul Surface layer Structure material AISI 304 Microstructure Layer compound Stainless steel has been used as structure material of power reactors of LWR, AGR and LMFBR types. The problem of structural materials that arises is low corrosion resistance and swelling in high irradiation environments. One method to overcome this problem is surface coating of the stainless steel cladding with chromium oxide. The aim of this study is to obtain the characteristics of microstructure, microhardness, and compounds formed in the surface layer of AISI 304 stainless steel. DC-sputtering method was used for the coating of AISI 304 surface. The coating process used coater material of chromium target and argon as sputtering gas which was mixed with oxygen gas with concentrations varying from 0 - 50% volume. The process was carried out at 10 - 20 mA current for 1 - 3 hours. Characterization of the layer includes microstructure observation using optical microscope, hardness with microhardness Vickers tester, and compounds analysis in layers with X-ray diffractometer. The results shows that the base metal microstructure is composed of austenite and carbide deposits. The metal-layer interface is clearly visible. The thickness of the resulted compact and homogeneous layer tends to increase for O2 gas composition in the range of 0 – 30 %volume but decrease for the range of 30 - 40 %volume and rise at 40 - 50 %volume. The hardness of the layer increases until oxygen gas is at 30 %volume and then decreases at 50 %volume of O2 gas. At a fixed value of O2 composition at a current of 10 - 20 mA and a process time of 1 - 3 hours, it is observed that the hardness of the SS 304 surface layer increases with increasing current and sputtering time. At a composition of 30 %volume of O2, 10 mA current and 2 hours process, the resulted layer contains CrO2. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7813 10.17146/urania.2019.25.2.5433 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 71-80 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7813/5958 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7814 2026-04-29T04:24:32Z urania:ART

OKSIDASI SIKLIK PADA TEMPERATUR 1200 oC PADUAN TERNER Ni-Al-Ti YANG DIDOPING DENGAN WOLFRAM Aprilianda, Adam Ambardi, Pradoto Prajitno, Djoko Hadi Ni base superalloy Cyclic oxidation Wolfram This research is aimed to study the effect of Wolfram (W) on high temperature oxidation resistance of Ni-Al-Ti ternary alloy by cyclic oxidation method at 1200 °C. The amount of Wolfram added was 1.5% W. Number of samples used in this study is 2 samples, namely Ni-Al-Ti alloy and Ni-Al-Ti-1.5W alloy. Cyclic oxidation process was carried out with cycle variation of 4, 9, 16 and 25 cycles (a cycle refers 60 minutes of heating in the furnace and 30 minutes cooling outside the furnace). The tests include Rockwell-C hardness testing, metallographic testing, oxide thickness testing, surface morphology observation, Scanning Electron Microscopy (SEM) examination and X-Ray Diffraction (XRD) analysis. Wolfram content shows an influence in the hardness values with the highest hardness value obtained for the Ni-Al-Ti-1,5W sample with a hardness of 48.5 HRC for as cast sample and 50.5 HRC after treatment solution. The addition of wolfram appears to improve the resistance to high temperature, which can be seen from the cyclic weight loss curve where the Ni-Al-Ti-1,5W sample has lower weight loss when compared to the other sample. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7814 10.17146/urania.2019.25.2.5442 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 83-92 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7814/5959 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7815 2026-04-29T04:24:32Z urania:ART

OPTIMASI PARAMETER PEMISAHAN Zr DALAM PEB U3Si2/Al PRA IRADIASI MENGGUNAKAN METODE KROMATOGRAFI PENUKAR ANION Yanlinastuti, Yanlinastuti Noviarty, Noviarty Haryati, Iis Fatimah, Syamsul Boybul, Boybul Ginting, Aslina Br. Separation Zirconium Fuel element plate of U3Si2/Al Anion exchange chromatography Determination of optimum parameters in the process of zirconium separation from U3Si2/Al fresh fuel solution by anion exchanger chromatography and analysis with UV-Vis spectrophotometer has been carried out. Zirconium is one of isotopes that can be used in the determination of fuel burn-up. The purpose of this study is to obtain optimal parameters of zirconium separation in standard materials as well as U3Si2/Al fresh fuel solution by anion exchanger chromatography. Optimization of Zr separation parameters was carried out using Dowex 1x-8 resin with a variation in weight of 0.5; 1.0; 1.5; 2.0 and 2.5 g, acidity of H2SO4 at 1; 2; 3; 4; 5 and 6 M, standard feed concentrations of Zr at 50; 100; 150 and 200 ppm and flow rate of 3M H2SO4 elution at 0.1; 0.3; 0.5; 0.7 and 0.9 mL/minute. To the Zr effluent as the result of separation was added Arsenazo III 0.1% as complexing agent, and subsequent analysis with UV-Vis spectrophotometry was done at a wavelength of 665.3 nm. The measurement results show that the optimal parameters of Zr separation are the use of 1.5 g resin with resulting recovery of 70.79%, acidity of 3M H2SO4 with recovery of 83.94%, feed concentration of 150 ppm with recovery of 87.64% at a flow rate of 0.5 mL/minute. When those optimal parameters were used in Zr separation from U3Si2/Al fresh fuel solution, the recovery obtained was 96.23%. These optimized parameters can also be used to separate Zr from U3Si2/Al irradiated fuel solution. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7815 10.17146/urania.2019.25.2.5508 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 93-102 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7815/5960 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7816 2026-04-29T04:24:32Z urania:ART

PEMISAHAN DAN ANALISIS CESIUM DALAM PEB U3Si2/Al TMU 2,96 gU/cm3 PASCA IRADIASI DENGAN METODE PENGENDAPAN CHLOROPLATINATE Nugroho, Arif Boybul, Boybul Indaryati, Sutri Haryati, Iis Kriswarini, Rosika Ginting, Aslina Br. U3Si2/Al Analysis of 137Cs isotopes Precipitation of Cs2PtCl6 Fuel burnup is an important parameter for determining the quality and integrity of irradiated nuclear fuel. The most important parameters in the calculation of burnup is the separation process and 137Cs content analysis sampled from the irradiated plate fuel. This research has conducted separation and analysis of the 137Cs contained in irradiated U3Si2/Al plate by chloroplatinate precipitation method. This method requires that the sample has to be made free from elements that may interfere with the precipitation process. The separation is done by precipitation of 137Cs isotope in the form of Cs2PtCl6. The precipitation process uses CsNO3 as non-active carrier with varied weight of 0.0263; 0.0501; 0.0754; 0.1036 ; 0.2064; 0.3020, and 0.4017 g. The CsNO3 carrier is added in order to obtain more ammount of precipitate of Cs in the form of Cs2PtCl6 because CsNO3 has similar chemical properties with 137Cs in uranyl nitrate solution. The 137Cs isotope is coprecipitated in Cs2PtCl6, and γ-spectrometry is used in the determination of the content of 137Cs in Cs2PtCl6 precipitate. The separation results show that addition of 0,1036 g CsNO3  has recovered 0.000328 g/g-sample of 137Cs from 0.000465 g/g-sample in irradiated U3Si2/Al solution, or a recovery of 70.4753%. Recovery value by this method is still lower when compared to other 137Cs separation methods due to longer steps of separation process, which may cause many errors in the sample preparation process. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7816 10.17146/urania.2019.25.2.5466 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 103-114 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7816/5961 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7817 2026-04-29T04:24:32Z urania:ART

PEMBUATAN STANDAR CESIUM DARI LARUTAN PEB U3Si2/Al PASCA IRADIASI DENSITAS 2,96 gU/cm3 DENGAN METODE PENGENDAPAN CsClO4 Ginting, Aslina Br. Boybul, Boybul Nugroho, Arif Yanlinastuti, Yanlinastuti Noviarty, Noviarty Wurdiyanto, Gatot Chandra, Hermawan Solution of U3Si2/Al Cesium Calibration Secondary standard Spectrometer-γ A considerable amount of irradiated nuclear fuel solutions exists in the hot cell area of Radiometallurgy Instalation (IRM). These solutions were generated from sampling activities for burnup test. The solutions cannot be disposed due to the present of 137Cs isotope and other heavy elements that have a long half time, but the solutions can be used as a secondary source of 137Cs standard for analysis purposes. Post-irradiation fuel analysis is always conducted by a valid method and with a calibrated spectrometer-γ. The standard used in the calibration of spectrometer-γ energy or activity is 137Cs isotope. There has been some restriction in the use of spectrometer-γ because spectrometer-γ calibration cannot be performed routinely due to the unavailability of standard sources. The availability of 137Cs isotope for R&D is still dependent on abroad suppliers. Therefore, in this research, PTBBN and PTKMR aim to develop a secondary standard source of 137Cs isotope with 10330 ± 411 Bq activity. The standard source was obtained from a solution of 137Cs separated from U3Si2/Al irradiated fuel. The separation was done by precipitation method with CsNO3 and HClO4 as the carrier. The result of the separation is 137CsClO4 precipitate, dried and weighed for obtaining greater activity and measured with a spectrometer-γ. The dried precipitate 137CsClO4 was subsequently packed as closed standard source (shield source).  This standard sourced was certified by PTKMR as secondary standard isotope of 137Cs and is ready to be used for energy calibration of spectrometer-γ. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7817 10.17146/urania.2019.25.2.5457 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 115-126 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7817/5962 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7820 2024-12-23T01:40:15Z urania:ART

KETAHANAN MULUR DAN OKSIDASI SIKLIK BAJA SUS 304 HASIL PELAPISAN METAL MATRIX COMPOSITE NiCr+Cr3C2 DENGAN METODA HVOF Naufal, Muhammad Luthfi Waldi, Muhammad Prajitno, Djoko Hadi SUS 304 HVOF Thermal Spray Coating Metal Matrix Composite Creep Resistance Cyclic Oxidation CREEP AND CYCLIC OXIDATION RESISTANCE OF SUS 304 STEEL AFTER COATING WITH METAL MATRIX COMPOSITE NiCr+Cr3C2 BY HVOF METHOD. Deposition of 60% NiCr + 30% Cr3C2 + 10% Al2O3 and 60% NiCr + 40% Cr3C2 metal matrix composite coating has been done on SUS 304 steel surface by the High Velocity Oxygen Fuel (HVOF) Thermal Spray Coating technique. The purpose of the coating was to improve SUS 304 steel creep resistance and cyclic oxidation on boiler tubes of steam power plant application. The treatments include creep test at 550oC with a load of 0.4 and 0.6 σy (yield strength) for 30240 minutes (21 days), cylclic oxidation with various temperatures of 600, 700 dan 800oC in tubes furnace for 1 hour and cooling for 15 minute in 1cycle. Measurement was done every one cycle to observe weight change of the oxide, and themeasurement was done for 5 cycles. The recommended coating specimen for boiler tubes of steam power plant application is coating specimen of 3 three components of MMC with the lowest creeprate of 0.000148917 %/hour and cyclic oxidation resistance at 600o (which has a lowest change inoxide weight, coating thickness and oxide compounds formed on coating). Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7820 10.17146/urania.2020.26.1.5693 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-12 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7820/5964 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7821 2026-04-29T04:24:32Z urania:ART

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL Dwijayanto, R. Andika Putra Husnayani, Ihda Zuhair RDE Spent fuel Thorium HTGR Tl-208 There are several options of nuclear fuel utilization in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilizes low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilize other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fueled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0.627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7821 10.17146/urania.2019.25.2.5525 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 127-140 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7821/5963 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7822 2024-12-23T01:41:44Z urania:ART

KETAHANAN OKSIDASI ZIRCALOY-4 YANG DIPADU DENGAN YTTRIUM PADA TEMPERATUR TINGGI Immanuel, Perdana Ambardi, Pradoto Prajitno, Djoko Hadi zircaloy-4 Isothermal oxidation Yttrium variation Yttrium is one of the rare earth metals which is used to improve properties of metals. This study was conducted to determine the effect of Yttrium (Y) on high-temperature oxidation resistance of zirconium alloy 4 by using isothermal oxidation method at temperatures of 900 °C and 1000 °C. This study employs 3 varied yttrium concentration of 0 wt%, 0.5 wt% and 1 wt% as doping material for zircalloy-4 (Zr, Sn, Fe, and Cr). The isothermal oxidation process was conducted at temperatures of 900 °C and 1000 °C for 9 hours heating of samples as cast in a furnace. Tests were done to the samples after oxidation process, which include metallographic testing, hardness testing, X-Ray Diffraction (XRD) testing and oxide thickness measurements. Test results show that the varying content of yttrium does not significantly affect the value of hardness, but there is a tendency that greater yttrium content decreases hardness value. The addition of yttrium affects the oxidation resistance at high temperatures, which can be seen from the graph of oxide thickness as well as the morphology of the oxide surface of each sample. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7822 10.17146/urania.2019.25.1.4663 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-8 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7822/5965 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7823 2024-12-23T01:41:44Z urania:ART

PERILAKU ELEKTROKIMIA BAJA TAHAN KARAT SS 316 DALAM MEDIA NANO FLUIDA Prajitno, Djoko Hadi Setiawan, Jan SS 316 Corrosion ZrO2 Nanofluid EIS method Corrosion is a common problem in many engineering metals and alloys. Electrochemical method is powerful instrument to use as tool to study corrosion behavior of metals and alloys in corrosive environment. The present study is to evaluate interaction between stainless steel SS 316 and nanofluid containing nano particles of ZrO2 as an additive to demineralized water by electrochemical method. Corrosion potential, Tafel polarization, and electrochemical impedance spectroscopy (EIS) of stainless steel SS 316 were performed in both demineralized water and demineralized water containing nano particle 0.05% ZrO2 at different immersion time. Corrosion potential examination shows that stainless steel 316 is actively corroded in both demin water and nanofluid. The corrosion potential tends to increase toward positive value for longer immersion time, both in demin water and nanofluid media. The value of anodic Tafel slope of also tends to increase for longer immersion time, both in demin water and nanofluid. Tafel polarization examination shows that corrosion rate of stainless steel 316 in demin water and nanofluid is not significantly different, and corrosion rate decreases with immersion time. EIS technique shows that the impedance of stainless steel 316 in nanofluid is higher when compared to the impedance in demineralized media, resulting in lower corrosion rates in nano fluids. X-ray diffraction examination of the specimens shows that the phase present in the stainless steel 316 alloys after corrosion testing is γ austenite, both in demin water and nanofluid. Stainless steel SS 316 is very good corrosion resistant material both in demin water and nanofluid with corrosion rate under 1 MPY. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7823 10.17146/urania.2019.25.1.5053 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 9-18 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7823/5966 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7824 2024-12-23T01:41:44Z urania:ART

ADSORPTION OF URANIUM SIMULATION WASTE USING BENTONITE: TITANIUM DIOXIDE Basuki, Kris Tri Hasnowo, Lutfi Aditya Jamayanti, Elza Bentonite TiO2 Adsorption Uranium Bentonite is a clay material of high surface area that have galleries within its structure. Bentonite that is modified with TiO2 will have high adsorption capability. In this study, natural bentonite and bentonite: TiO2 were characterized with FTIR, XRD and BET instruments to determine functional group, basal spacing, and specific surface area. This study also investigates the adsorption of bentonite:TiO2 in various environmental factors, such as pH (pH 1, 3, 5, and 8), contact time (10, 20, 30, 40, 50, 60, 70, 90, and 120 min), and initial uranium concentration (20, 40, 60, 80 ppm), and their influences on adsorption capacity, and determine the kinetics equation and adsorption isotherm. Based on FTIR analysis, a decrease in the band of O-H bond from water molecule was observed, which indicates the presence of TiO2 in bentonite interlayer structure. The XRD characterization of bentonite: TiO2 does not show diffraction peak in 001 plane. This is due to delamination of bentonite interlayer structure. Delamination is caused by the presence of TiO2 in large quantities, thus damaging the bentonite interlayer structure into irregular sheets. Bentonite as sheets will cause the basal spacing to increase and it is anticipated that XRD will find it difficult in detecting the 001 plane at a low 2 theta angle. The surface area of bentonite: TiO2 has increased by 12.04 m2/g. The maximum adsorption capacity of U(VI) took place at pH 5.0 for 70 minutes contact time and uranium concentration of 60 ppm. In this study, the adsorption kinetic and adsorption isotherm are pseudo second-order kinetic and Langmuir isotherm. The kinetic constant and maximum adsorption capacity of bentonite: TiO2 are 0.075 g/mg.min and 5.848 mg/g respectively. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7824 10.17146/urania.2019.25.1.4527 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 19-32 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7824/5967 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7825 2024-12-23T01:41:44Z urania:ART

GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE-BED REACTOR Adrial, Hery Hamzah, Amir Hartini, Entin Gamma dose rate HTGR 10 MWth Biological shield Pebble HTGR-10 MWth is a high-temperature gas-cooled reactor. The fuel and moderator are pebble shaped with a radius of 3 cm. One fuel pebble consists of thousands of UO2 kernels with a density of 10.4 gram/cc and the enrichment rate of 17%. The core of HTGR-10 MWth is the center of origin of neutrons and gamma radiation resulting from the interaction of neutrons with pebble fuel, moderator and biological shield. The various types of radiations generated from such nuclear reactions should be monitored to ensure the safety of radiation workers. This research was conducted using MCNP-6 Program package with the aim to calculate and analyze gamma radiation dose in biological shield of HTGR-10 MWth. In this study, the biological shield is divided into 10 equal segments. The first step of the research is to benchmark the created program against the critical height of HTR-10. The results of the benchmarking show an error rate of ± 1.1327%, while the critical core height of HTGR 10 MWth for the ratio of pebble fuel and pebble moderator (F:M) of 52:48 occurs at a height of 134 cm. The rate of gamma dose at the core is 3.0052E+05 mSv/hr. On the biological shield made of regular concrete with a density of 2.3 grams/cc, the rate of gamma dose decreases according to an equation y = 0.0042 e-0.03x. Referring to Perka Bapeten no 4 of 2013, the safe limits for workers and radiation protection officers will be achieved if the minimum thickness of biological shield is 115 cm with gamma dose rate of 0 mSv/hour. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7825 10.17146/urania.2019.25.1.4923 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 33-44 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7825/5968 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7826 2024-12-23T01:41:44Z urania:ART

KARAKTERISASI RADIONUKLIDA PADA BAHAN BAKAR BEKAS DARI EXPERIMENTAL PEBBLE BED REACTOR Aisyah, Aisyah Mirawaty, Mirawaty Saputra, Dwi Luhur Ibnu Setiawan, Risdiyana Characterization of radionuclides AVR Spent fuel TRISO coated pebble Arbeitsgemeinschaft Versuchsreaktor (AVR) is a type of High Temperature Gas Cooled Reactor (HTGR) that has a TRISO coated pebble fuel similar to TRISO fuel for the Experimental Power Reactor (Reaktor Daya Experimental, RDE), which is planned to be built in Indonesia. Therefore, the radionuclides characterization study in the AVR spent fuel can give contribution to the study of RDE spent fuel. One of the important aspects in the operation of nuclear reactor is the management of the generated spent fuel. The management of the AVR spent fuel was carried out by storing the spent fuel in dry cask for a long time. In designing the safety of spent fuel storage system, one of the needed important studies is the characterization of radionuclides contained in the spent fuel. In this study, characterization of radionuclides contained in the spent fuel has been performed by using ORIGEN 2.1 computer program based on the data of AVR reactor operation. The objective of the research is to analyze the safety of a long-term storage of pebble spent fuels in dry cask. The results show that up to 100 years storage period, one pebble spent fuel has a total activity concentration of 4.03x1010 Bq/g of activated products, actinides and its daughter and fission product radionuclides. In a dry cask containing 1900 pebbles, the total activity concentration of the spent fuel is 7.66x1013 Bq/g. Some broken TRISO layer in some spent fuel pebbles in dry cask is evident, so it is possible that some amount of fission product radionuclides is released from the spent fuel in the dry cask such as 85Kr, 135Xe, and 131I in the form of gas, and metallic radionuclides of 137Cs,106Ru, 110mAg and 107Pd. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7826 10.17146/urania.2019.25.1.5024 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 45-58 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7826/5969 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7827 2024-12-23T01:41:44Z urania:ART

SENSITIVITY ANALYSIS ON THERMOHYDRAULIC CODE FOR MODIFIED PLATE-FUELED 2 MW TRIGA Hastuti, Endiah Puji Widodo, Surip Sensitivity analysis TRIGA plate COOLOD-N2 Heathyd PARET-ANL The plan to modify TRIGA 2000 Bandung from using regular TRIGA fuel to plate-type fuel should be supported by the use of appropriate computer codes. This research proposes three codes to design reactor thermohydraulic at transient condition. Analysis has been performed to identify code sensitivity using the same input and correlation. The codes used were COOLOD-N2, Heathyd, and PARET-ANL. The input was obtained from preliminary analysis of a flow rate calculation of 70 kg/s and a nominal power of 2 MW. The comparison of these three codes did not consider uncertainty factor for neutronic and technical aspects. The sensitivity analysis on thermohydraulic codes used to calculate heat transfer in the fuel plate of TRIGA reactor at steady state condition indicates similar temperature trend lines for the coolant, plate, and fuel meat. Temperature calculation results obtained from COOLOD-N2, Heathyd and PARET ANL give consistent sensitivity with the differences of coolant temperature from 2.83% to 12.5%; cladding temperature  from 2.14% to 31.30%; and fuel meat temperature from 6.63% to 18.64%. The margins of flow instability were 5.03; 5.68 and 4.21, respectively for COOLOD-N2, Heathyd, and PARET-ANL. These values show that flow instability has not yet occurred. The results of the analysis show that the use of those three codes for steady state condition using the same input, in which uncertainty factor is neglected, give similar trend for coolant, cladding, and fuel meat temperature. As the modelling in each code is different, the values obtained are not exactly the same. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7827 10.17146/urania.2019.25.1.4961 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 59-70 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7827/5970 Copyright (c) 2019 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7832 2024-12-23T01:40:15Z urania:ART

PENGARUH METODE PREPARASI BROTH DAN JENIS PREKURSOR ZIRKONIUM PADA KARAKTERISTIK KERNEL YTTRIA-STABILIZED ZIRCONIA (YSZ) Sarjono, Sarjono Mutiara, Etty Yusnitha, Erilia Dewayatna , Winter YSZ kernel sol-gel preneutralization urea decomposition external gelation TRISO THE INFLUENCE OF SOL PREPARATION METHOD AND TYPE OF ZIRCONIUM PRECURSOR ON THE CHARACTERISTIC OF YTTRIA-STABILIZED ZIRCONIA (YSZ) KERNEL. Fabrication of yttria-stabilized zirconia (YSZ) kernel by external gelation method has been done. Urea was used in the broth preparation to serve as: 1) as homogenous precipitation agent to produce sol when added at decomposing temperature of urea and 2) as pore promoter when added at below decomposing temperatures of urea. The research used urea in varied mol urea/metal ratio to obtain an the optimum value based on the resulted morphological properties of the sintered kernel. The zirconium precursor used was zirconium oxynitrate hydrate. Characterization of the kernel products include examination with SEM to study the morphological properties of the dried and sintered kernels, measurement of diameter and density of the sintered kernel, and XRD examination to determine the phase of the resulted solid solution of zirconium-yttrium oxide. Based on themorphological examination, the optimized addition of urea according to its function: 1) as precipitation agent in the sol preparation, mol urea/metal ratio of 1.0 at decomposing temperature of 80oC and 2) as porosity generation in the gel microsphere, mol urea/metal ratio of 1.5 at addition temperature of 50oC. The XRD examination shows that a mixture of 16% mol yttrium nitrate dan 84% mol zirkonium nitrate has resulted in solid solution of Y0,15Zr0,85O1,93 or 92ZrO2.8Y2O3 kernels with cubic phase. Increase in precursors concentration in broth has resulted in increase in the diameter of the sintered YSZ kernel, where kernels with average diameter of 0,510 mm was obtained from 0,855 mol/L precursors in broth. Kernels with desired diameter can be prepared by tuning of precursors concentration in broth. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7832 10.17146/urania.2020.26.1.5743 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 13-24 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7832/5973 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7839 2024-12-23T01:40:15Z urania:ART

PENGARUH JENIS PACKING PADA MENARA PACKED-BED ABSORBER DALAM PENYERAPAN GAS NOx Kadarjono, Agoeng Yusnitha, Erilia Santosa, Agus Sartono Dwi Winastri, Pertiwi Diah HTGR kernel gas effluent packed-bed packing design THE EFFECT OF PACKING TYPE ON THE PACKED-BED ABSORBER IN ABSORPTION OF NOx. Design of NOx effluent gas processing needs to be done for broth preparation and gel casting activities in the UO2 kernel fabrication process series. The availability of this effluent processing tool will increase the economic value and process efficiency, reduce the  amount of the chemicals needed, and minimize the hazards associated with toxicity and contamination. The design of  the effluent processing tool chose the packed-bed absorber using NOx gas as a feed stream and water as an absorbent medium. The packed-bed absorber has been designed and calculated to determine the effect of various types of packing (pall, ralu, and nor-pac ring) for the absorbtion of the NOx gas. To determine the effect of the packing type, it is carried out by simulating fluid flow rate on diameter, effective surface area of packing, pressure drop, mass transfer coefficient, packing column height, and tower height by inserting into the appropriate equations using the EXCEL program. The inclusion of the effect of feed flow rate and type of packing in the calculation resulted in smaller diameter for the tower containing pall ring packing than the diameter of tower containing other fillings but gave higher pressure drop values, effective surface area of packing, total mass transfer coefficient, and tower height. Increase in NOx gas absorption causes an increase in the amount of packing space and tower height. Increase in operating pressure and temperature causes a decrease in packing and tower height, whereas increase in flooding factor, beside causing a decrease in packing and tower height, decreases the tower diameter. This type of pall ring packing is a design choice because it produces small/economic dimensions. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7839 10.17146/urania.2020.26.1.5738 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 25-36 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7839/5979 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7841 2024-12-23T01:40:15Z urania:ART

PENENTUAN RECOVERY NEODYMIUM DAN CERIUM PADA PROSES PEMISAHAN ISOTOP DARI SIMULASI PEB U3Si2/Al PRA IRADIASI Noviarty, Noviarty Noerpitasari, Erlina Kriswarini, Rosika Jamaludin, Agus Indaryati, Sutri Nisa, Sayyidatun Fatimah , Samsul Nd recovery Ce rcovery Burn-up Indicator U3Si2/Al DETERMINATION OF NEODYMIUM AND CERIUM RECOVERY IN ISOTOPE SEPARATION PROCESS OF SIMULATED PRE-IRADIATED U3Si2/Al FUEL. Neodymium (Nd) is one of the fission product isotopes used as an indicator of nuclear fuel  burn-up because Nd isotope is a stable fission product, has a sufficiently great yield, is not volatile and has low neutron-capture cross- section. Burn-up determination using Nd can support and verify the results of the burn-up determined by  using Cs, U and Pu isotopes. Destructive burn-up determination (Destructive Test, DT) is based on the determination of  specific isotopes that is performed physicochemically using mass spectrometer equipment. Based on ASTM E 321-69  method for Nd isotope analysis, Nd separation from Ce is required because 142Ce isotope interferes with 142Nd  measurement due to their being isobars, and this phenomenon will affect the results of the burn-up calculations. In  addition, Nd isotope in U3Si2/Al fuel is present along with other fission isotopes such as 235U, 239Pu, 148Nd, 137Cs,  152Eu, 90Sr, 143Ce, 103Ru, 95Zr, and 95Nb. Therefore, to obtain an accurate value to be used as burn-up indicator, the  isotopes must be separated. Before the Nd and Ce separation, several separation phases of Cs isotopes (137Cs, 134Cs), U isotopes (234U, 235U, 236U, 238U), Pu isotopes (239Pu, 238Pu) and other isotopes (152Eu, 90Sr, 103Ru, 95Zr and 95Nb) were done. The determination of separation recovery aimed to determine the amount of Nd and Ce at each stage of the  separation process of other isotopes prior to the separation stage of Nd and Ce. The determination of recovery was  conducted by a simulation of fresh U3Si2/Al solution plus standard solutions of Cs, U, Nd and Ce. The recovery for Nd  and Ce from Cs separation phase using zeolites were 69.91% and 64.42% respectively while the recovery of Nd and Ce at  Pu and U separation phase were close to 100%. The recovery values of Nd and Ce were then used as a correction in the  calculation of Nd and Ce separation from the fuel solution, and the corrected values can then be used as a reference in determining the burn-up of U3Si2/Al fuel using Nd as an indicator. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7841 10.17146/urania.2020.26.1.5791 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 37-48 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7841/5982 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7844 2024-12-23T01:40:15Z urania:ART

UJI TAK MERUSAK PELAT ELEMEN BAKAR U3SI2/Al DENSITAS URANIUM 4,8 gU/cm3 MENGGUNAKAN RADIOGRAFI SINAR-X DIGITAL Artika, Refa Sigit, Rohmad Rahmatullah, Helmi Fauzi Supardjo, Supardjo Ginting , Aslina Br. Fuel plate U3Si2/Al 4,8 gU/cm 3 non destructive test X-ray radiography NON DESTRUCTIVE TEST FOR U3Si2/Al FUEL PLATE WITH 4,8 gU/cm3 URANIUM DENSITY USING DIGITAL X-RAY RADIOGRAPHY. Non-destructive test for irradiated U3Si2/Al fuel plate with 4,8 gU/cm3 uranium density and 20, 40, and  60% burn-up has been done. The test includes visual inspection, radiographic image analysis, and dimension measurement using digital x-ray radiography. The non-destructive test aimed to determine the performance, integrity, and reliability of U3Si2/Al fuel plate with 4,8 gU/cm3 uranium density during irradiation at RSG-GAS. Reliability and safety of the fuel during irradiation are indicated by the absence of anomalies in the post-irradiation examination result. Three samples of irradiated U3Si2/Al fuel element (FE) with 20, 40, and 60 % burn-up were transferred from RSG-GAS to RMI hot cell 101 through transfer channel of Interim Storage Spent Fuel (ISSF). The U3Si2/Al fuel plate were then transferred to hot cell 103 for non-destructive test using digital x-ray radiography with seven shots for good resolution radiographic images. The radiographic images were evaluated to obtain data of defects and dimensions such as meat width, fuel plate width, meat-cladding distance, right-left meat- cladding boundary distance at the far side (SJ) and the near side  (SD). The X-ray radiography analysis shows the presence of white spot in the middle of U3Si2/Al fuel cladding with 20% burn- up. Meanwhile, indication of surface defects and other defects were not found for the 40 and 60% burn-up samples. The dimensional measurement data show no dimensional changes, which indicates solid integrity after irradiation in the RSG-GAS. The measurement data analisis can further be used for mapping the sampling spots for the next test. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7844 10.17146/urania.2020.26.1.5822 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 49-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7844/5986 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7848 2024-12-23T01:40:15Z urania:ART

PROLIFERATION RESISTANCE ASSESSMENT OF INDONESIAN 10-MWT RDE EXPERIMENTAL POWER REACTOR USING INPRO METHODOLOGY Farawan, Bening Husein, Muhammad Sukron Fajrin Sunarto, Agus Adhi , Arief Sasongko Proliferation Resistance Safeguards NPRO Methodology PROLIFERATION RESISTANCE ASSESSMENT OF INDONESIAN 10-MWT RDE EXPERIMENTAL POWER REACTOR USING INPRO METHODOLOGY. Assessment of proliferation resistance (PR) for 10-MWt RDE has been conducted to verify the RDE safeguards desain. Proliferation resistance (PR) is a parameter that can be used to measure the ability of a facility to implement safeguards (safeguardability). Safeguards is every technical measure used to ensure that every nuclear material in a facility is used as declared and only for welfare. Proliferation resistance of 10-MWt RDE has been assessed with INPRO methodology. The areas assessed were specific to intrinsic features of RDE, which consist of the attractiveness of nuclear material and technology and also detectability and chances of diversion of the nuclear material. This assessment concludes that the 10-MWt RDE has strong proliferation resistance, which indicates that the facility has small ability to diverse nuclear material and will less likely be used to proliferate nuclear weapon. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7848 10.17146/urania.2020.26.1.5620 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-68 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7848/5988 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7856 2024-12-23T01:42:39Z urania:ART

PRODUCTION OF URANIUM-MOLYBDENUM ALLOY AS A CANDIDATE FOR NUCLEAR RESEARCH REACTOR FUEL Suryaman, Ganisa Kurniati Wildan, Muhammad Waziz Supardjo Susanto, Yatno Dwi Agus U−Mo alloy research reactor Fuel Research and development on high density uranium for nuclear research reactor fuel is still in progress. Uranium-molybdenum alloy is one of the strongest candidates of nuclear research reactor fuel material. The properties and characteristics of U-Mo alloy is of important consideration for the selction of the fabrication techniques for the production of  the fuel. In this work, uranium-molybdenum (U-Mo) alloys with varied molybdenum content have been produced succesfully by arc melting technique. The molybdenum content variations were 7 %wt, 8 %wt, 9 %wt and 10 %wt Mo. The melting process was done 5 times to achieve homogenization. Metallographic micrograph shows the presence of dendritic structure. XRD examination result affirms the presence of 2 phases of γ-U phase and d-U2Mo phase. Microhardness Vickers test shows higher hardness value for Uranium-molybdenum alloy with higher molybdenum content. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7856 10.17146/urania.2018.24.3.4685 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 135-142 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7856/5993 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7857 2024-12-23T01:42:39Z urania:ART

SIMULASI DIAMETER GEL BASAH PADA FABRIKASI KERNEL YTTRIA-STABILIZED ZIRCONIA MENGGUNAKAN ALAT GEL-CASTING Yusnitha, Erilia Sarjono Susilowati, Sri Rinanti Dewayatna, Winter Sediawan, Wahyudi Budi Simulation Diameter Microsphere of gelled kernels Broth Gel-casting equipment In kernel fabrication, microsphere of gelled kernels (yttria-stabilized zirconia) were made by droplet formation of the broth through the gel-casting equipment. Broth is feed of gel-casting equipment that were made of mixed of zirconium (IV) nitrate, yttrium (III) nitrate hexahydrate, urea, tetrahydrofurfuryl alcohol (THFA), and poly vinyl alcohol (PVA). The process parameter of the gel-casting equipment such as vibration frequency and broth flowrate need to be set up to control shape and diameter of the targeted microsphere of gelled kernels. The gel-casting equipment at PTBBN – BATAN is equipped with single nozzle with 1 mm diameter. The simulation work is important to be carried out to reduce the number of experiments in the laboratory hence it reduces the waste from trial and error in the experiment. Moreover, the objective of this simulation work is to predict the diameter of microsphere of gelled kernels. Therefore, the simulation of microsphere of gelled kernel diameter is necessary to be build and verified with the experiment data. According to hypothesis, the simulation of microsphere of gelled kernel diameter can be conducted by considering gel-casting equipment parameter such as the broth flowrate and vibration frequency. In addition, the broth characteristic as broth density affects the diameter of microsphere of gelled kernels diameter. The diameter of microsphere of gelled kernels produced from gel-casting equipment were determined with digital microscope. The result from simulation calculation shows that diameter of microsphere of gelled kernels produced from laboratory experiment can be approached by modified equation involving vibration frequency, broth flowrate, metal concentration in the broth and broth density. It appears that the deviation of simulation calculation is smaller than simple equation, where only vibration frequency and broth flowrate are included in the calculation. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7857 10.17146/urania.2018.24.3.4984 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 143-154 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7857/5994 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7858 2024-12-23T01:42:39Z urania:ART

PREDIKSI SIFAT MATERIAL PRESSURE TRANSMITING MEDIUM PADA QUASI-ISOSTATIC PRESSING MENGGUNAKAN FINITE ELEMENT ANALYSIS Sutarya, Dede Pamungkas, Iqbal Ramadhian Sukma, Hendri Sartono, Agus Haryadi, Dedi Quasi-isostatic pressing Pebble fuel Mooney-Rivlin Pressure transmitting medium Finite element analysis   Quasi-isostatic pressing technique is developed from isostatic pressing, which is made more simple and efficient. Quasi-isostatic pressing can provide the effect of volumetric triaxial compaction, so that it can produce compaction with uniform density. With these advantages, the quasi-isostatic pressing technique is used in pebble fuel fabrication process for high-temperature gas-cooled reactors (HTGR). The main problem in the quasi-isostatic pressing technique lies in need of hyperelastic material as a pressure transmitting medium (PTM) to transmit uniform static pressure in all directions in the compacting process. Therefore, the properties and characteristics of the PTM material needed for quasi-isostatic pressing with the boundary conditions of process must be determined. In this paper Finite Element Analysis (FEA) was used to predict the properties and characteristics (constants) of PTM materials for the quasi-isostatic pressing process using Mooney-Rivlin 2 and 3 parameters and single-acting press methods. From the results of FEA experiments, C10 values of 120 MPa, C01 of 49 MPa and C11 of 10 MPa were obtained. However, the quasi-isostatic pressing process for graphite powder compacting has not completed perfectly, so further research will be carried out with numerical improvement and the use of the Mooney-Rivlin 5 to 9 parameter model using double-acting press method. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7858 10.17146/urania.2018.24.3.5028 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 155-166 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7858/5995 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7859 2024-12-23T01:42:39Z urania:ART

ANALISIS POTENSI PROSES CREEP (PEMULURAN) TEHADAP BAHAN BAKAR UJI AKIBAT DISTRIBUSI STRAIN (KETEGANGAN) DENGAN MENGGUNAKAN PROGRAM FEMAXI Yulianto, Tri Wigayanti, Etty Marti Nuclear fuel elements Pin Cladding Pellets Analysis of creep process potential of test fuel due to strain distribution has been carried out using FEMAXI program. The purpose of this study is to predict the possibility of creep of test fuel due to circumocular distribution of axial and radial strain at certain operating temperatures and times and to provide information on quality standards during production and operation of pin tests in the reactor core. This analysis is important because cladding is a barrier to the pressure release of fission gas products produced when the fuel is irradiated. The phenomenon of the total creep process on pin cladding is assumed to be caused by the effect of burn-up on the release of fission gas products, which results in pressure induced by strain of cladding at relatively high temperatures. The creep rate will increase with increasing strain. The total mechanical analysis of the top  and inner  part of fuel pins at steady operating conditions shows that there is an increase in creep depending on the LHR (linear heat rate) and operating pattern on the reactor core. For transient operating conditions, it shows an increase in stable creep. In ramp operating conditions or sudden LHR change operations, it indicates an increase in creep processes due to difference in strain pressures between steady operation patterns, transients and ramps during the process of creep on pellets and cladding. This is due to fission gas products, but the final results of the total pin condition calculation indicate that the pin products are still considered as safe in terms of both thermal and mechanical processes with various patterns of LHR changes. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7859 10.17146/urania.2018.24.3.5031 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 167-176 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7859/5996 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7860 2024-12-23T01:42:39Z urania:ART

EVALUATION OF RADIATION DOSE RATE OF RSG-GAS REACTOR Hamzah, Amir Adrial, Hery Subiharto Dose rates RSG-GAS Radiation safety MCNP The RSG-GAS reactor has been operated for 30 years. Since the nuclear reactor has been operated for a long time, aging process on its components may occur. One important parameter for maintaining the safety level of the RSG-GAS reactor is to maintain radiation exposure as low as possible, especially in the working area. The evaluation results should be able to demonstrate that the radiation exposure of the RSG-GAS is still safe for workers, communities and the surrounding environments. The purpose of this study is to evaluate radiation exposure in the working area to ensure that the operation of RSG-GAS is still safe for the next 10 years. The scope of this work is confirming the calculation results with the measured radiation dose in the RSG-GAS reactor working area. Measurement of radiation exposure is done by using the installed equipments at some points in the RSG-GAS working area and a portable radiation exposure measurement equipment. The calculations include performance of a modeling and analysis of dose rate distribution based on the composition and geometry data of RSG-GAS by using MCNP.  The analysis results show that the maximum dose rate at Level 0 m working area of RSG-GAS reactor is 3.0 mSv/h with a deviation of 6%, which is relatively close to the measurement value. The evaluation results show that the dose rate in RSG-GAS working area is below the limit value established by the Nuclear Energy Regulatory Agency of Indonesia (BAPETEN) of 10 mSv/h (for the average effective dose of 20 mSv/year). Therefore, it is concluded that the dose rate in RSG-GAS working area is safe for personnel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7860 10.17146/urania.2018.24.3.4909 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 177-186 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7860/5997 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7861 2024-12-23T01:42:39Z urania:ART

PENGARUH PH PADA PENURUNAN KADAR THORIUM DALAM LIMBAH MENGGUNAKAN PROSES ELEKTROKOAGULASI DENGAN ELEKTRODA ALUMUNIUM DAN TEMBAGA Prayitno Ridantami, Vemi Mulyani , Indah Muji Mulyani Electrocoagulation pH Coagulant Reduction of thorium level by electrocoagulation process has been done. Electrocoagulation process is a method used in waste management, which is based on the use of electrodes with potential difference to produce coagulant. One of the process parameters in electrocoagulation is pH. This study aims to determine the effect of pH on the electrocoagulation process in order to lower the thorium level. The method used was batch electrocoagulation process in electrocoagulator tub of 22.5 × 14.5 × 20 cm dimension. The electrodes used were aluminum and copper of 20 × 20 cm2 size with active surface area of 250 cm2. Electrodes were installed in parallel with 1 cm distance between the electrodes. The electrocoagulation process was carried out with a working voltage of 2 V for 60 minutes. Sampling was done every 5-minute interval to prepare samples for subsequent thorium level analysis and pH measurement. The results show changes in pH value during the electrocoagulation process. Such changes result in the concentration change, which was indicated by coagulant separation efficiency. The separation efficiency of aluminum electrode was 98.06%, which was indicated by an increase in thorium level   from the initial value of 439.274 ppm to 8.503 ppm within 5 minutes at a pH of 7.38. The copper electrode shows 96.35% efficiency in reducing thorium level from the initial value of 439.274 ppm to 16.015 ppm within 5 minutes at a pH of 7.29. Both electrodes used in the electrocoagulation processes show high separation efficiency and significant increase in pH values within 5 minutes, suggesting that the formation rate of coagulants is rapid during 5 minutes. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7861 10.17146/urania.2018.24.3.4396 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 187-198 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7861/5998 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7863 2024-12-23T01:43:08Z urania:ART

PENGARUH TEMPERATUR DAN WAKTU HOMOGENISASI TERHADAP KARAKTERISTIK PELAT PADUAN AG3NE Hadijaya Aidah, Siti Masrukan AG3NE AlMgSi Homogenization Characterization Structural material Investigation on the effects of homogenization temperature and time on AG3NE alloy plate has been done. AG3NE alloy is used as cladding material of research reactor fuel. Ingot of AlMgFeSi melt was sampled for several tests and examination such as chemical composition, hardness, microstructure, phase, density and corrosion. Composition analysis shows that the alloy is composed of Al= 96.110%, Fe= 0.188%, Mg= 2.333% and Si= 0.355%.The hardness value for AG3NE is 153.33 HV for samples treated with rapid cooling and 72.70 HV for samples treated with slow cooling. Microstructure examination with optical microscope shows that AG3NE alloy without homogenization consists of flat grains with a fraction of dendrite of 0.02 mm mean diameter, while samples homogenized at 500°C at varied temperatures of 1, 2, 3, and 4 hours consist of grains with 0.01 mm, 0.02 mm, 0.03 mm and 0.04 mm respectively. Phase analysis with X-Ray Diffractometer indicates the existence of Al (ά) and FeSiAl5 phases at 2θ of 38.24° - hkl (119) and 44.446° - hkl (209) with 100% and 76.6% intensity respectively. The measured density of AG3NE, however, is similar to that of AlMgSi as material for comparison, the values of which are 2.6893 ± 0.0013 g/cm3 for AG3NE and 2.6910 ± 0.0283 g/cm3 for AlMgSi. Corrosion rate analysis shows that the corrosion rate for AG3NE is lower than that for AlMgSi. Based on the characterization for chemical composition, microhardness, microstructure, density and corrosion rate, it is concluded that AG3NE alloy can be used as a candidate for cladding material for research reactor fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7863 10.17146/urania.2018.24.2.4391 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 73-84 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7863/6000 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7865 2024-12-23T01:43:08Z urania:ART

KARAKTERISTIK MEKANIK KELONGSONG BAHAN BAKAR NUKLIR TEMPERATUR TINGGI DALAM ATMOSFER GAS NITROGEN Sigit, Rohmad Ismarwanti, Sri Guswardani Susanto, Yatno Dwi Agus Setiawan, Jan Mechanical properties Cladding Zircaloy Nitrogen Nitriding At Loss of Cooling Accident (LOCA) scenario, Zr-2 and Zr-4 as nuclear fuel cladding potentially exposed to air which experimentally leads to mechanical properties degradation faster than in steam environment. This study aims to determine the impact of nitrogen on the mechanical strength of cladding by minimizing other environmental influences. The Zr-2 and Zr-4 cladding was annealed under vacuum at a temperature of 500°C for one hour. It was followed by nitriding with the pressure of 1500 mbar for two hours. It was followed by natural cooling to room temperature. The results of tensile strength and microhardness testing showed that tensile strength, ductility and microhardness values increase of 12.1%, 19.2% and 28.2% for Zr-2 and 5.1%, 25% and 22.6% for Zr-4, respectively. Nitride layer was formed on the cladding surface due to interaction with nitrogen and there is no mechanical properties degradation due to nitriding process. It was caused by nitriding processes that were not followed or accompanied by oxidation processes. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7865 10.17146/urania.2018.24.2.4422 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 85-92 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7865/6001 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7866 2024-12-23T01:43:08Z urania:ART

CORE DESIGN OF TRIGA2000 BANDUNG USING U3Si2/Al FUEL ELEMENT MTR TYPE Pinem, Surian Surbakti, Tukiran Sembiring, Tagor Malem Cylinder fuel type MTR fuel type Batan-FUEL Thermal neutron flux The TRIGA2000 Bandung reactor currently uses cylinders fuel type, but the fuel production has been discontinued. In order to operate continuously, it is planned that the TRIGA2000 reactor should change the fuel from cylinder type to U3Si2/Al MTR type since Indonesia has been able to produce such fuel element. In this research calculation of TRIGA2000 coreusing U3Si2/Al fuel MTR type of three different fuel densities has been done. This activity begins with performing neutron macroscopic cross section generation for all core materials as a function of temperature, burn up and xenon. The cross-section generation was performed with the WIMSD5 code. The calculation of the core parameters were done with the Batan-FUEL code. Based on the calculation of neutronic parameters, there are three core configurations, i.e.core with 16 fuel elements and 4 control elements (Core 16/4), core with 14 fuel elements and 4 control elements (Core 14/4) and core with 12 fuel elements and 4 control elements (Core 12/4). The three cores meet the safety constraint but only the Core 16/4 that uses U3Si2/Al fuel with a density of 2.96 gU/cm3. The maximum thermal neutron flux in the core center is 5.874×1013 n/cm2s with core cycle length is 245 days at 2 MW. The results show that the TRIGA2000 core can be converted from cylinder type fuel to silicide MTR type. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7866 10.17146/urania.2018.24.2.4302 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 93-104 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7866/6002 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7867 2024-12-23T01:43:08Z urania:ART

ANALISIS KOROSI PIPA PENDINGIN SEKUNDER RSG-GAS DENGAN TEKNIK ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY (EIS) Ajiriyanto, Maman Kartaman Kriswarini, Rosika Yanlinastuti, Yanlinastuti Lestari, Diyah Erlina Corrosion Secondary cooling pipe EIS Inhibitor Multi-purpose reactor G.A. Siwabessy has a secondary cooling system facility to extract heat generated from uranium fission reactions in the reactor. The cooling pipe material is a type of carbon steel. Secondary cooling media is water supplied from Puspiptek. To extend the lifetime of the pipe, chemical compound ZnPO4 was put on the water-cooling media as a corrosion inhibitor. The corrosion analysis of secondary pipe was performed by corrosion testing using EIS and tafel method. The electrochemically corrosion test was performed with three electrode cells in water media, 0.05 M HCl solution and 3% NaCl solution. The aim is to obtain the optimum inhibitor concentration added in the secondary coolant and to determine the inhibition mechanical as well as the inhibitor efficiency. EIS testing was performed in the frequency range 100 kHz to 0.02 Hz in HCl 0,05M. Test results show changes in electrical parameters such as Rct, Cdl and Rs. The inhibitor concentration increases then Rct value increase. At inhibitor concentration of 90 ppm obtained the highest Rct value is 81,29 ohm. Efficiency of the inhibitor addition produces maximum value is 66,75% at concentration of 90 ppm. The corrosion rate in samples on medium of 3% NaCl without inhibitor, moss inhibitor and corrosion inhibitor were 3.975, 3.576 and 3.215 mpy respectively. The addition of a mixture of corrosion inhibitors and moss inhibitors can decrease the corrosion rate up to 1.586 mpy quantitatively. The addition of inhibitors using ordinary water may inhibit interface reactions on the surface of the sample as indicated by the higher the inhibitor concentration the higher the Rct value. In NaCl 3% medium also shows the same phenomenon relatively that the corrosion mechanism can be inhibited by increasing of Rct value as well as in this condition, Rs value is constant relatively. The dominant fact that the addition of a corrosion inhibitor could decrease the corrosion rate. The moss inhibitors have a primary function not to decrease the corrosion rate. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7867 10.17146/urania.2018.24.2.4421 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 105-114 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7867/6004 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7868 2024-12-23T01:43:08Z urania:ART

OPTIMASI PH ALUMINA DAN 99Mo DALAM PEMBUATAN GENERATOR 99Mo/99mTc BERBASIS MoO3 ALAM Sulaiman Sugiharto, Yono Chairuman Setiawan, Gatot Gunawan, Adang Hardi 99Mo Alumina Coloumn 99Mo pH Alumina material has long been used for 99Mo radioisotope adsorbent in the 99Mo/99mTc generator. The 99Mo/99mTc generator produces 99mTc radioisotope which is used for the diagnostic purposein nuclear medicine. To date, alumina is used to adsorp 99Mo obtained from fission reaction of 235U. The restriction of the use of fission 235U leads to the find of an alternative 99Mo. The alternative of 99Mo production is obtained from natural MoO3 which is irradiated by thermal neutron in the nuclear reactor. The 99Mo radioisotope produced from this reaction has no carrier-free properties. This is due to the abundanceof natural MoO3 is 24.13%.In the99Mo/99mTc generator based on natural MoO3 by using alumina column, it is necessary to know the alumina type for optimum 99Mo adsorption, as well as pH of 99Mo. In this study, there were used three variations of alumina-based generator column namely acid, neutral, and basic alumina. The 99Mo solution used in this study was also performed three pH: pH 4, pH 7, and pH 8 of the 99Mo solution. The generator eluted and observed daily for 3 days and continued after the next two days for 2 days. The optimum adsorption of 99Mo from natural MoO3  by alumina material was given in the use of acidic alumina with the pH 4 of the99Mo solution. Among the three pH variation of the 99Mo solution, the 99Mo solution of pH 4 gives the highest of 99mTc yield. The column with the smallest 99Mo breakthrough is given by acidic alumina column using the 99Mo solution of pH 4. Thus, acidic alumina column can be used as an adsorbent to adsorp natural 99Mo. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7868 10.17146/urania.2018.24.2.4159 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 115-124 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7868/6005 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7869 2024-12-23T01:43:08Z urania:ART

PEMISAHAN CESIUM DALAM PEB U3Si2/Al PASCA IRADIASI DENGAN METODE KOLOM PENUKAR KATION MENGGUNAKAN RESIN DOWEX Nugroho, Arif Anggraini, Dian Boybul, Boybul Indaryati, Sutri Kriswarini, Rosika Ginting, Aslina Br. U3Si2/Al Isotope137Cs Dowex resin cation exchange column burn up The burn up calculations of nuclear fuel by destruction method can be determined from the analysis of 137Cs isotope, as fission product. For this reason, the separation of 137Cs isotopes from uranium in irradiated fuel element plates of U3Si2/Al (PEB) needs to be done. The separation of 137Cs was done by using cation exchange column method with Dowex resin with weight variation of 1; 2; 3; 4; 5 and 6 g. This study aims to determine the content of 137Cs in irradiated fuel element plates of U3Si2/Al (PEB) using the cation exchange column method with Dowex resin to be compared with the result of 137Cs separation by cation exchange column method using Lampung zeolite and CsClO4 precipitation method. The irradiated fuel element plates of U3Si2/Al (PEB) PEB U3Si2/Al with a density of 2.96 gU/cm3 from the middle part of RI-SIE2 code and side plate 20 IDA 0045 was taken with weight 0.1103 g and dissolved with HCl 6N and HNO3 6N to obtain uranil nitrate solution. The U3Si2/Al post irradiated solution was 150 μL was added with 2 mL aquadest and added with Cs carrier 20 μL and 1 mL HCl 12 M. The solution mixture was used as feed into the anion exchange column (first column) containing the R-Cl resin –with the wight variation about 1; 2; 3; 4; 5 and 6 g. The effluent from the anion exchange column was introduced into the cation exchange column (second column) which contained R-NH4+ resin with weight variation of 1; 2; 3; 4; 5 and 6 g. The bounded 137Cs isotope in R-NH4+ resin in the column was being eluted using 1 M HCl of 15 mL and being vaporized until the volume reach ± 2 mL. The quantity of the 137Cs isotope in the effluent was measured by using a gamma spectrometer with a counting time of 500 s. The 137Cs isotope content in irradiated fuel element plates of U3Si2/Al with the density 2.96 gU/cm3is 0.000445 g/gPEB and the parameter of separation 137Cs with cation exchange column method using Dowex resin weighing 2 g with recovery is 98.67%. The result of separation of 137Cs by cation exchange column method is not much different from the result of separation 137Cs by cation exchange method and CsClO4 precipitation method. The 137Cs obtained can then be used for burn up calculations. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7869 10.17146/urania.2018.24.2.4306 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 125-134 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7869/6006 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7870 2024-12-23T01:39:49Z urania:ART

PENGARUH ROL DINGIN DAN ANIL TERHADAP KARAKTERISTIK STRUKTUR MIKRO DAN KEKERASAN MIKRO PADUAN Zr-0,5%Nb-Si Hardiyanti, Heri Adhi, Arief Sasongko Setiawan, Jan cold-roll recrystallization microstructure microhardness Zr-Nb-Si alloy EFFECTS OF COLD-ROLLING AND ANNEALING ON THE MICROSTRUCTURE AND MICROHARDNESS CHARACTERISTICS OF Zr-0.5%Nb-Si. The performance improvement of zirconium alloys can be done by adding a certain percentage of common alloying elements such as Nb and Si. The addition of Nb provides resistance to irradiation and stabilizes the β phase (which normally forms at high temperatures) at room temperature. The addition of Si increases the strength of the alloys. The synthesis of a Zr-0.5% Nb-Si alloy have been carried out using an electric arc furnace. The microstructure and the microhardness of this alloy were observed after cold-rolling and annealing. It can be seen from the annealed alloy without cold-rolling that the grain size is larger with reduced microhardnes. The alloy treated with cold-rolling shows changes in the microstructure while the deformation density increases, which results in an increase in microhardness. Heat treatment to the cold-rolled alloy changes the microstructure of the alloy where recrystallization occurs, and the deformation density decreases, which results in the decreasing of microhardness. The lowest grain size and microhardness for the alloy that was cold-rolled at ε = 0.8 and annealed are 0.302 mm and 187 HVN, respectively. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7870 10.17146/urania.2020.26.2.6069 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 75-82 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7870/6007 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7871 2024-12-23T01:39:49Z urania:ART

THE CONCEPTUAL DESIGN FOR LIQUID EFFLUENT TREATMENT OF UO2 KERNEL FABRICATION Yusnitha, Erilia Kadarjono, Agoeng Sartono, Agus Winastri, Pertiwi Diah liquid effluent UO2 kernel CHEMCAD HTGR THE CONCEPTUAL DESIGN FOR LIQUID EFFLUENT TREATMENT OF UO2 KERNEL FABRICATION. The pebble fuel for HTGR is prepared through fabrication of UO2 kernel, coated particle and spherical element fuel. In the fabrication of UO2 kernel by external gelation method, a multicomponent of liquid effluent is generated. Therefore, the liquid effluent is required to be treated for safety reason before disposed to waste storage. In this paper, the conceptual design for theliquid effluent treatment of UO2 kernel fabrication is performed with the simulation process using CHEMCAD software. CHEMCAD is a software that can be utilized for chemical process design. The results showed that the proposed conceptual design is able to separate valuable components: isopropyl alcohol (IPA) and tetrahydrofurfuryl alcohol (THFA). The flowrate of IPA product is 5.28 kg/h with purity of 0.99 in mass fraction and the flowrate of THFA product is 1.01 kg/h with purity of 0.99 in mass fraction. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7871 10.17146/urania.2020.26.2.5881 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 69-74 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7871/6008 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7874 2024-12-23T01:39:49Z urania:ART

UJI TAK MERUSAK SHORT PIN PWR – FUEL DUMMY PASCA IRADIASI Sigit, Rohmad Artika, Refa Ajiriyanto, Maman Kartaman Yulianto, Tri short pin PWR cladding post-irradiation non destructive test X-ray radiography NON-DESTRUCTIVE TEST OF IRRADIATED SHORT PIN OF PWR TYPE – FUEL DUMMY. Post- irradiation examination (PIE) for nuclear fuel is used as an indicator of fuel performance reliability during the irradiation process. The Center for Nuclear Fuel Technology (PTBBN) conducts research and development of nuclear fuel for Pressurized Water Reactor (PWR). Short pin of PWR fuel dummy with zircaloy-4 cladding has been irradiated at the Multipurpose Reactor – G.A. Siwabessy (RSG-GAS) and has been sent to the Radiometallurgy Installation (IRM) – PTBBN. Non-destructive tests conducted at the IRM includes visual observation and X-ray radiographic test. Visual observations using a digital camera were done in hot cell 102, while X-ray radiographic test was carried out in hot cell 103 IRM. X-ray radiographic test using the voltage and current parameters at a range of 120-125 kV and 1000-1100 uA. The test material was placed at a height of 60 cm from the detector to obtain a 100% magnification. Visual observation shows the formation of several spots indicating oxidation on the short pin PWR fuel dummy surface. Radiographic images from X- ray radiographic testing do not show any anomalies, which is indicated by relatively homogeneous gray value in the axial and radial directions. Generally, based on the non-destructive test results, performance of the short pin PWR - fuel dummy is relatively satisfactory with no significant defects. To confirm the existence of oxides formation on the surface of the short pin PWR fuel dummy, Eddy Current test is needed, followed by microstructure observations. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7874 10.17146/urania.2020.26.2.6058 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 83-90 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7874/6035 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7876 2024-12-23T01:43:28Z urania:ART

STRATEGY TO SUPPORT HTGR FUEL FOR THE 10 MW INDONESIA’S EXPERIMENTAL POWER REACTOR (RDE) Taryo, Taswanda Ridwan Sunaryo, Geni Rina Rachmawati, Meniek RDE Indonesia HTGR Strategy Fuel The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power (25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and roadmap for the preparation of the RDE fuel plant construction with the involvement of national stakeholders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7876 10.17146/urania.2018.24.1.3729 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-16 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7876/6011 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7877 2024-12-23T01:43:28Z urania:ART

MIKROSTRUKTUR DAN KOMPOSISI FASA PELET SINTER UO2 DENGAN PENAMBAHAN DOPAN LOGAM ZIRKONIUM Yulianto, Tri Mutiara, Etty UO2 pellet Zr PWR Microstructure Phase Characterization of zirconium-doped UO2 pellet for PWR fuel has been done. The investigation aims to study the influence of zirconium doping on microstructure, phase composition and stoichiometric (O/U) ratio. Homogenous distribution of zirconium metal is expected in order to enhance UO2 pellet quality. The UO2 pellet was fabricated by compaction and sintering with varied zirconium addition of 0.3, 0.5 and 0.9 wt%. Sintering was done at 1700 °C with heating rate of 250 °C/hour for 3 hours in hydrogen medium. Characterization of the sintered pellet includes visual examination, density measurement, microstructure examination with optical microscope at 500x magnification and phase analysis by X-ray diffraction (XRD) method. Analysis results show that the addition of varied Zr of 0.3, 0.5 and 0.9 wt% did not cause defect or crack of the sintered UO2 pellet. Microstructure examination with optical microscope, however, could not visually confirm the existence of zirconium. The evaluation of XRD data was done with HighScore software by conforming peaks data of the test sample with peaks data of a reference database. The phase of UO2 pellet without zirconium addition was confirmed to be UO2 and UO3. The phase of UO2 pellet with zirconium addition, on the other hand, shows that the UO3 fraction increases linearly with the percentage of zirconium added, which is indicated by the value of stoichiometric (O/U) ratio. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7877 10.17146/urania.2018.24.1.3941 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 17-26 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7877/6012 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7878 2024-12-23T01:43:28Z urania:ART

KARAKTERISASI PADUAN AlMgSi UNTUK KELONGSONG BAHAN BAKAR U3Si2/Al DENSITAS URANIUM 5,2 gU/cm3 Ginting, Aslina Br. Supardjo Yanlinastuti, Yanlinastuti Anggraini, Dian Boybul, Boybul U3Si2/Al Density 5,2 gU/cm3 AlMgSi cladding AlMg2 Increase in uranium density from 2.96 gU/cm3 to 5.2 gU/cm3 of U3Si2/Al fuel requires a compatible cladding. High density fuel possesses high hardness value, and the use of AlMg2 as cladding material may cause dog bone during rolling process. In addition to dog bone phenomena, swelling can also occur during the use of nuclear fuel in the reactor due to rising in fission products and burn up. For this reason, substitution of cladding material should be determined to suit the high density U3Si2/Al fuel. This research deals with the characterization of AlMgSi as a candidate for AlMg2 substitution. The characterization includes thermal analysis with DTA (Differential Thermal Analysis) dan DSC (Differential Scanning Calorimetry), hardness test with macrohardness tester, microstructure examination with SEM (Scanning Electron Microscope), and corrosion analysis by heating at 150 °C for 77 hours in an autoclave. Analysis results show that both AlMgSi and AlMg2 share stable heat compatibility with U3Si2/Al up to 650 °C. AlMgSi cladding has a hardness value of 115 HVN, higher than that of AlMg2, which is 70.1 HVN. Microstructure examination shows that AlMgSi has better morphology of interface bonding than AlMg2. Corrosion rate analysis also indicates that AlMgSi has lower corrosion rate. The results of thermal, hardness, microstructure and corrosion characterization reveal that the U3Si2/Al fuel of 5.2 gU/cm3 suits better with AlMgSi than AlMg2. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7878 10.17146/urania.2018.24.1.4016 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 27-38 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7878/6013 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7879 2024-12-23T01:43:28Z urania:ART

ANALISIS METALOGRAFI PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 2,96 gU/cm3 PASCA IRADIASI Ajiriyanto, Maman Kartaman Ginting, Aslina Br. Junaedi PEB U3Si2/Al Post-irradiation Microstructure Cladding thickness Metallographic analysis of U3Si2/Al fuel element plate has been performed in hot cell. The purpose of metallographic analysis is to study changes in PEB U3Si2/Al microstructure and AlMg2 cladding thickness after irradiation in reactor until burn up of 56%. The fuel element plate of irradiated U3Si2/Al was cut in top, middle and bottom positions with each size around 5x5x1,37mm. Metallographic preparation starts from sample cutting using cutting machine with low speed and sample mounting, grinding and polishing in hot cell 104–105. Sample mounting was done by using resin for more than 10 hours followed by grinding with sandpapers up to grit size of 2400 and polishing with diamond paste of size 3 to 1 micron at a rotational speed of 150 rpm for 5 minutes. Microstructure observation was performed with optical microscope in hot cell 107 at 200 times magnification. Microstructure examination reveals U3Si2 particles with diverse forms and sizes, Al matrix and AlMg2 cladding were spread along the U3Si2/Al side. Microstructure observation of irradiated U3Si2/Al has not shown good result because only topography observation of U3Si2/Al meat, Al matrix and AlMg2 cladding can be done due to limited capability of the optical microscope in hot cell, where maximum magnification can be attained only at 200 times so that the phenomenon of interaction layer and small gas bubble cannot be observed. However, U3Si2/Al microstructure of 56% burnup, if compared to the microstructure of U3Si2/Al fuel element plate of 60% burnup from previous researcher, shows interaction between U3Si2 meat with Al matrix and the existence of layers with a thickness about 5 up to 20 microns. Meanwhile, the observed thickness of AlMg2 cladding is greater than 0.25 mm, which indicates that irradiation does not significantly change the thickness of AlMg2 cladding so that the overall irradiated U3Si2-Al still has good integrity and stability. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7879 10.17146/urania.2018.24.1.4010 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 39-50 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7879/6014 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7880 2024-12-23T01:43:28Z urania:ART

ANALISIS PROBABILISTIC FRACTURE MECHANICS PADA EVALUASI KEANDALAN BEJANA TEKAN REAKTOR SECARA 3-D Hartini, Entin Himawan, Roziq Susmikanti, Mike Probabilistic fracture mechanics Pressure vessels 3-D Material integrity analysis is required for the reactor pressure vessel (RPV). The component is the pressure boundary that serves to contain radioactive materials. The presence of cracks on the walls can affect the integrity of the RPV. The objective of this study is to analyze fracture mechanics using probabilistic models for the evaluation of RPV reliability. Probabilistic models are used as an analysis approach for random character of input quantities such as the mechanical properties of materials and the physical environment. The random character of input quantities was obtained by sampling technique base on probability density function. The material used in this study was ferritic steel (SA 533). Fracture mechanics analysis was performed by Finite Element Method (FEM) with MSC MARC software. The outputs of MSC MARC, which is of J integral values, were converted into stress intensity factor (SIF) for evaluating the reliability of RPV in 3D model. It is found that the calculation resulted from the SIF by probabilistic method reaches a limit value of fracture toughness earlier than the SIF by deterministic method. The SIF generated by probabilistic method is 95,8 MPa m0.5, while the SIF generated by deterministic method is 91,8 MPa m0.5, where lower ratio value of (a/c) will produce greater value of SIF. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7880 10.17146/urania.2018.24.1.3721 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 51-60 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7880/6015 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7881 2024-12-23T01:43:28Z urania:ART

DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL Rahayu, Dyah Sulistyani Purwanto, Yuli Salimin, Zainus Heat and radiation decay Spent fuel Storage cask The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decay at a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectively. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method based on SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640; 335.2; and 298.551 watts. The result of the trial-and-error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7881 10.17146/urania.2018.24.1.3694 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 61-72 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7881/6016 Copyright (c) 2018 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7882 2024-12-23T01:37:33Z urania:ART

PENGARUH TEMPERATUR DAN KOMPOSISI Nb TERHADAP INTERDIFUSI PADUAN U-Zr-Nb Ajiriyanto, Maman Kartaman Masrukan Al Afghani, Fajar Yanlinastuti, Yanlinastuti Intermic U-Zr-Nb alloy Aluminum Microstructure Hardness Phase A simulation of U-Zr-Nb intermixed fuel has been carried out with aluminum plates. Initially, U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb alloy ingots were made using an electric arc melting furnace. The ingot obtained was cut into pieces with a thickness of 2-3 mm and a diameter of about 5-6 mm, then it was covered Al plate on one side and pressed. After pressing, the alloy ingots of U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb were heated at temperatures of 350 and 450 oC for 1 hour. The ingot was subjected to several tests, namely microstructure examination using optical microscope and Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS), hardness test using microhardness tester, and phase observation using X-ray diffractometer (XRD). The results of microstructure examination show that the boundaries between the U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb layers (interfaces) and the Al layer were clearly visible. The higher of the Nb content, the denser the interface between the two layers. For the microhardness test, we found that the hardness at the interface was difficult to measure, while the U-Zr-Nb region shows an increase in hardness for heating from 350 to 450 oC, while in the Al section there is a decrease in hardness when the heating temperature increases from 350 to 450 oC. UO3 compounds were formed in all test samples while the γ-U phase was formed in U-6Zr-2Nb samples both heated at temperatures of 350 and 450 oC. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7882 10.17146/urania.2021.27.3.6503 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 111-122 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7882/6017 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7883 2024-12-23T01:37:33Z urania:ART

PERILAKU TARIK DAN STRUKTUR MIKRO BAJA TAHAN KARAT AISI 304 PASCA PERLAKUAN PANAS PADA DAERAH SENSITISASI 600 – 700 °C Sungkono Ismarwanti, Sri Stainless steel Sensitization area Tensile strength Hardness Microstructure Precipitate AISI 304 stainless steel is used as a structural component of the LWR type power reactors. In the manufacture of structural component, the quality of SS 304 welded joints is determined by cooling process from the austenite to sensitization temperature. The purposes of this study were to determine tensile strength, hardness and microstructure of SS 304 in the sensitization area of 600 – 700 °C. The research method was testing of the chemical composition, tensile, hardness and microstructural properties referring to ASTM. The results show that the tensile strength of SS 304 after sensitization treatment at 600 °C (30-60 minutes) and 700 °C (30-90 minutes) was relatively the same and lower than samples without heat treatment while the sample treated at 600 °C (90 minutes) was relatively the same as the sample without heat treatment. The microstructure of SS 304 consists of d-ferrite and austenite phases and chromium carbide precipitates. The quantity of precipitate at the grain boundaries decreases with increasing temperature and holding time. The hardness after sensitization treatment at 600 – 700 °C was relatively the same on the base metal, near the fracture and the fracture area. The most crack resistant was obtained on SS 304 samples with treatment at 700 °C (90 minutes). Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7883 10.17146/urania.2021.27.3.6570 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 123-132 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7883/6018 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7884 2024-12-23T01:37:33Z urania:ART

ANALISIS KETIDAKPASTIAN PENGUKURAN KEKUATAN TARIK DAN ELONGASI SPESIMEN SS304 BERBENTUK RING Ismarwanti, Sri Setiawan, Jan Ajiriyanto, Maman Kartaman Sigit, Rohmad Uncertainty GUM Tensile strength Elongation SS304 ring Uncertainty is a measure of the reliability of a measurement. One of the widely used uncertainty analysis approaches is the Guide to the expression of uncertainty in measurement (GUM). The steps for analyzing the uncertainty of a measurement have been described in detail by the GUM approach. Uncertainty analysis was carried out on tensile testing to determine the traceability of the data obtained. Tensile testing produces data in the form of tensile strength and elongation of the test material. Tensile strength is a quantity that involves the magnitude of the force and length. Elongation is the increase in length when pulling. The tensile strength of the ring-shaped sample will be affected by the cross-sectional area on both sides of the ring when a tensile stress is applied. So, to determine the uncertainty of the tensile test, an analysis of the contributors to the uncertainty is required. This paper provides an explanation of how to analyze the uncertainty of the measurement of tensile strength and elongation in a ring-shaped sample. In addition, the value of the uncertainty in the measurement of tensile strength and elongation of SS304 (in the form of a ring) will also be presented. The test results show that the average tensile strength of SS304 rings is 631.070 ± 26.976 Nmm-2 with an expanded uncertainty of 50.451 Nmm-2 at a 95% confidence level. In addition, at the same level of confidence, the average SS304 ring elongation result is 18.681% ± 2.425% with an expanded uncertainty of 2.867%. Data repetition or repeatability is the biggest contributor to the uncertainty in measuring the tensile strength and elongation of the SS304 ring. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7884 10.17146/urania.2021.27.3.6567 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 133-142 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7884/6019 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7886 2024-12-23T01:37:33Z urania:ART

OPTIMASI PARAMETER METODE PEMISAHAN RADIONUKLIDA 95Zr DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al TERIRADIASI Yanlinastuti, Yanlinastuti Noviarty, Noviarty Haryati, Iis Saputri, Ariyanti Boybul, Boybul Supardjo Ion exchange column irradiated U3Si2/Al 95Zr radionuclide γ-spectrometer Horwitz test Radionuclide of 95Zr is one of the γ-radiation emitter radionuclides which has a short half-life so that it can be used as an indicator of burn up. In the irradiated U3Si2/Al fuel element plate there are several radionuclides of fission products emitting α, β and γ and when measuring radiation using γ-spectrometry, these radionuclides may interfere with each other resulting in inaccurate measurement results. Therefore, it is necessary to separate 95Zr for better accuracy in the measurement of γ-radiation with a γ-spectrometer. The purpose of this study was to obtain the optimum parameters for the separation of 95Zr in U3Si2/Al fuel element plate solution by ion exchange column method using Dowex 1x-8Cl- resin of 100-200 mesh in diameter. The optimum parameters obtained were then used for the separation of 95Zr in a solution of irradiated U3Si2/Al fuel element. It was found that the optimum parameter of acidity of H2SO4 in the feed was a concentration of 0.5 M, and the acidity of H2SO4 for elution was 1.0 M with a recovery of 98.20%. The weight of the resin to bind 95Zr radionuclide was 2.0 g with a recovery of 78.76%. The required feed volume was 50 µL with a recovery of 96.90% and the optimized flow rate was 0.1 mL/minute with a recovery of 96.72%. The optimum values were used for the separation of the U3Si2/Al fuel element plate solution irradiated with the code of Bottom (B), Middle (M) and Top (T). The measurement results show that the mean recovery of separation was 95Zr for code B-1 = 75.078%; B-2= 81.401%; M1=76,850%; M-2=83.806%; T-1=84.433%; and T-2 = 81.728% with the acceptability of the CV repeatability value smaller than the CV Horwitz value. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7886 10.17146/urania.2021.27.3.6535 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 143-156 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7886/6020 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7887 2024-12-23T01:37:33Z urania:ART

PEMISAHAN CESIUM DALAM BAHAN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 BURN UP 60% POTONGAN BOTTOM Saputri, Ariyanti Indaryati, Sutri Haryati, Iis Yanlinastuti, Yanlinastuti Mahmudah, Hanifah Rifaatul Ginting, Aslina Br. U3Si2/Al density 4,8 gU/cm3 Burn up 60% Cesium Separation The separation of cesium in U3Si2/Al fuel with a density of 4.8 gU/cm3 and 60% burn up by cation exchange using zeolite from Lampung has been carried out. Cesium isotope is used as a burn-up indicator for irradiated nuclear fuel. The separation aims to obtain an accurate cesium weight so that the calculation of burn up by the destructive method can be carried out correctly. The fuel plate of U3Si2/Al was cut in duplicates at the bottom with a weight of B-1=0.036 g and B-2= 0.037g, respectively. The fuel plate of U3Si2/Al with this weight was then dissolved using 6N HCl and 6N HNO3 in 10 mL. The solution of U3Si2/Al was then pipetted 1 mL and transferred from the hot cell to R.135 and diluted to 25 mL. The solution was pipetted 50 µL into a vial in duplicates and 1000 mg of Lampung zeolite was added. Separation process of 134Cs and 137Cs from 235U was done by cation exchange for 60 minutes. The results obtained were isotopes 134Cs and 137Cs in the solid phase while uranium (238U, 235U, 234U) and other isotopes in the liquid phase. The weight of the isotopes of 134Cs and 137Cs was then measured using a spectrometer/g. The weight of the isotopes of 137Cs and 134Cs in solution at Bottom position was B-1=0.00003283 g/0.036g and B-1= 0.000000147 g/0.036 g for 137Cs and 134Cs respectively while another duplicate was B-2=0.00003290 g/0.037g for 137Cs and B-2=0.000000222 g/0.037g for 134Cs. It can be concluded that in the separation of cesium in 50 µL solution of U3Si2/Al at Bottom position by cation exchange using 1000 mg of zeolite from Lampung, good results were obtained. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7887 10.17146/urania.2021.27.3.6571 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 157-164 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7887/6021 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7888 2024-12-23T01:37:33Z urania:ART

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY Zuhair Luthfi, Wahid Dwijayanto, R. Andika Putra Suwoto Criticality Safety parameters NuScale fuel assembly MCNP6 ENDF/B-VII NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7888 10.17146/urania.2021.27.3.6509 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 165-176 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7888/6022 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7889 2024-12-23T01:44:47Z urania:ART

IDENTIFIKASI FASA PADA PELET BAHAN BAKAR U-ZrHx HASIL PROSES SINTER DENGAN ATMOSFER NITROGEN Masrukan Setiawan, Jan Biyantoro, Dwi Sintering U-ZrHx pellet Phase Atmosphere Nitrogen Phases identification of U-ZrHX  fuel pellets sintered under nitrogen atmosphere has been carried out. This sintering process is intended to obtain higher density pellets that meet the requirements for use as a fuel in pressurized water reactor (PWR). The results of this study are expected to be used for the determination of phase and density changes occurred in U-ZrHx pellets at various Zr concentrations after sintering at various temperatures. PWR fuel from U-ZrHx pellet needs to be sintered to increase its density. At first, pellets made from U-ZrHx powder with Zr composition of 35, 45 and 55 wt.% through a pressing process at a pressure of 509-637 MPa. The resulted U-ZrHx green pellets were subsequently sintered under nitrogen gas atmosphere at 1200 °C and 1300 °C for a fixed sintering time of one hour.  Phase analysis of the sintered U-ZrHx pellets was done by X-ray diffraction (XRD) technique and the data obtained were processed by using  HighScore Plus software. The test results with XRD indicate the formation of phases such as U(ZrHr2), UO3, UO2, and Zr. The UO2 and Zr phases were characterized for their density, lattice parameter and percentage of weight. In this experiment it is desired to obtain U-ZrHx with high density without the formation of many phases. Density measurement shows that the density of zirconium metal and UO2 decreases with increasing sintering temperature (from 1200 °C to 1300 °C), while the lattice parameter increases with increasing sintering temperature. Data calculation of weight percentage for sintering at 1300 °C shows that the percentage values of the resulted zirconium metal and uranium dioxide are the closest to their theoretical calculation results. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7889 10.17146/urania.2017.23.3.3641 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 139-152 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7889/6023 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7890 2024-12-23T01:44:47Z urania:ART

ANALISIS CACAT BLISTER PADA KELONGSONG BAHAN BAKAR U3Si2/Al MENGGUNAKAN ULTRASONIC TEST Fauzi, Muhammad Himawan, Roziq Rahmatullah, Helmi Fauzi Permana, Setia Gogo, Antonio Ultrasonic test U3Si2/Al fuel cladding Blister Pulse-echo Through transmission Fuel integrity is one of safety considerations during nuclear reactor operation. Any defect resulted after reactor operation should be avoided. Thus, in nuclear fuel development, post-irradiation examination is necessary to be performed to ensure fuel integrity. Non-destructive test using ultrasonic method is one of alternative methods to perform post-irradiation examination. The objective of this analysis is to detect blister defect in cladding of plate type nuclear fuel. In post irradiation examination (PIE) using ultrasonic method, the operation parameters are necessary to determine. In this experiment, this analysis method was conducted on U3Si2/Al fuel element cladding consisting of blister defect. Ultrasonic tests were done using two techniques, i. e. pulse-echo and through transmission techniques. The results were presented in A-scan display. The results show that the longitudinal wave speed in water used in the testing is 1545 m/s. The optimal distance between the probe and the specimen when using pulse-echo technique is 20 mm. The optimal distance between the two probes when using through transmission technique is 40 mm. Tests at blister area using pulse-echo technique caused output signal to decrease by 19 %FSH, while through transmission technique caused output signal to decrease by 80 %FSH. According to the results it can be concluded that ultrasonic test is effective to detect a blister defect in plate type nuclear fuel, and through transmission technique exhibited a clearer difference between normal area and blister area in fuel cladding when compared to pulse-echo technique. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7890 10.17146/urania.2017.23.3.3629 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 153-164 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7890/6024 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7891 2024-12-23T01:44:47Z urania:ART

KARAKTERISTIK DAKTILITAS SS304 YANG TEROKSIDASI PADA TEMPERATUR TINGGI Setiawan, Jan Sungkono Ductility SS304 Oxidation High temperature This paper shows changes in hardness, tensile strength and ductility characteristics of SS304. The SS304 was processed by heat treatment at high temperature in atmospheric environment. SS304 is a type of stainless steel with many superior properties used in industry. In this research, a pipe made of SS304 with 30 mm in diameter was cut in transversal direction at 3 mm width to be used as a sample.  The samples were processed by heat treatment at 600, 800 and 1000 °C for 1 and 2 hours in atmospheric environment.  After the oxidation process, the samples were cooled by natural cooling outside the furnace or quenched in water.  Hardness testing and ring tensile testing were done to the high-temperature oxidized SS304.  Analyses were done to study changes in hardness, tensile strength and ductility characteristics of SS304 after oxidation process.  The results show that hardness escalation occurs mainly for samples that were cooled by quenching in water compared to natural cooling. The highest hardness value was observed for the sample oxidized at 1000 °C for 1 hour and cooled by quenching, where hardness value changed from 281.25 HV before oxidation to 468.00 HV after oxidation.  The tensile strength of the samples, however, tend to decrease, while the ductility of some samples tends to increase.  Increase in tensile strength followed by increase in ductility occurs for samples oxidized at 600 °C for 2 hours with natural cooling.  The tensile strength after oxidation reaches a value of 744.04 MPa with 12.38 % elongation escalation. As a guidance for implementation, SS304 in high temperature oxidation environment should be limited by temperature of 600 °C. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7891 10.17146/urania.2017.23.3.3645 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 165-174 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7891/6025 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7892 2024-12-23T01:44:47Z urania:ART

PENGARUH HIDROGEN TERHADAP KEKUATAN TARIK KELONGSONG BAHAN BAKAR NUKLIR BERBASIS PADUAN ZIRKONIUM Sigit, Rohmad Al Afghani, Fajar Setiawan, Jan Sungkono Hydrogen Tensile strength Cladding Ring tensile test Ductility During reactor operation, zircaloy-4 (Zr-4) as fuel cladding is under an environment that can potentially cause mechanical failure, especially by the effect of cladding-hydrogen interaction. Hydrogen would diffuse and precipitate to form brittle hydrides when the solubility limit is surpassed. In this experiment, ring tensile test method was used to examine the effect of hydrides existence on tensile strength of Zr-4 cladding. The Zr-4 sample was annealed at 500 °C and 600 °C for an hour in vacuum condition and followed by hydriding with a pressure up to 1000 mbar with 3 hours holding time and slow cooling. According to ideal gas equation, hydrogen content in Zr-4 reaches about 674 ppm and 1215 ppm at hydriding temperature of 500 °C and 600 °C respectively. Microstructure observation shows that hydrides precipitate was formed at grain boundaries. The hydrides were found to be longer and thicker at higher hydriding temperature. As the hydrogen content increases, the ring tensile test shows increasing stress at 4.4 %, decreasing strain at 2.1 % and increasing microhardness level at 13 %. This result confirms that the existence of hydrides precipitates in cladding has an effect of increasing the tensile strength and decreasing the ductility obviously. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7892 10.17146/urania.2017.23.3.3648 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 175-182 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7892/6026 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7893 2024-12-23T01:44:47Z urania:ART

ANALISIS KOROSI PADUAN ZIRLO-Mo DALAM MEDIA NaCl MENGGUNAKAN METODE POLARISASI Ajiriyanto, Maman Kartaman Anggraini, Dian Kriswarini, Rosika Corrosion rate Polarization method Zirlo-Mo alloy Zirlo-Mo is a zirconium type of alloy, which is developed as cladding material for light water reactor. As cladding material, the important parameter to be characterized is its corrosion resistance properties. The goal of this research is to study the corrosion resistance of zirlo-Mo in NaCl medium with varied concentration of 0.03; 0.3; and 3 % and compare it to the those of zircaloy-2 dan zircaloy-4 in the same medium. Samples were mounted into disk form and metallographically prepared, i.e. treatment with abrasive paper of 500 to 1200 grit SiC. The analysis of zirlo-Mo corrosion rate was performed by polarization method with a potentiostat. The advantage of this polarization resistance method is that it takes shorter testing time if compared to weight gain method, and it can be applied to analysis at low corrosion rate of below 0.1 mpy. Other tests were also done to support the corrosion rate analysis such as microhardness test with microhardness tester and microstructure examination with optical microscope. The tests results show that in general zirlo-Mo, zircaloy-2 and zircaloy-4 possess very low corrossion rate, i.e. below 1 mpy. The corrosion rate of zircaloy-2, zircaloy-4 and zirlo-Mo in NaCl 3 % medium are 0.006; 0.007 and 0.590 mpy respectively. The corrosion rate of zirlo-Mo is relatively the highest due to its lath martensite or accicular structure, which has high hardness value. The lath martensite structure and high hardness properties make zirlo-Mo highly susceptible to corrosion. Hardness test results show that the hardness value of zircaloy-2, zircaloy-4 and zirlo-Mo are 194.79; 227.95 and 360.05 HV respectively. The existence of solid solution of Mo and Zr-α as well as the martensite or acicular structure of zirlo-Mo result in high hardness value of the alloy. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7893 10.17146/urania.2017.23.3.3221 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 183-194 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7893/6027 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7894 2024-12-23T01:44:47Z urania:ART

PENGARUH PENAMBAHAN ION SULFAT TERHADAP REKOVERI PEMISAHAN URANIUM MENGGUNAKAN METODE KOLOM PENUKAR ANION Boybul, Boybul Yanlinastuti, Yanlinastuti Anggraini, Dian Ginting, Aslina Br. Nugroho, Arif Kriswarini, Rosika Separation Uranium Recovery HCL Na2SO4 solvents In the previous research, it was observed that the recovery of uranium separation by anion exchange column method with HCl 12 N solvent was 68.2137 %. The recovery can be increased by optimizing the uranium separation parameters by the use of HCl and Na2SO4 solvent mixture. A 500 μL uranyl nitrate standard solution was used as feed with varied HCl concentration of 6 N, 9 N, 12 N and varied Na2SO4 concentration of 500, 1000, 1500, 2000, 2500 and 3500 ppm as reagent solution. In the anion exchange process, uranium was bound by the resin to form [UO2(SO4)3]4- and [UO2(SO4)2]2 complex compounds, and these complexes were eluted with 15 mL of 0.15 N HCl dilute solution. The eluate was heated to nearly dried condition and 1 mL of HCl concentrated solution was added for electrodeposition process. The uranium deposited on stainless steel planset, as the result of electrodeposition process, was analyzed with alpha spectrometer. It was observed that the largest 235U content was found in 9M of HCl solvent and 1500 ppm of Na2SO4 with 95.7124 % recovery. Therefore, the concentration value of HCl of 9 N and 1500 ppm of Na2SO4 were subsequently applied to the uranium separation of irradiated U3Si2/Al fuel element plate. An amount of 500 μL supernatant of U3Si2/Al specimen taken from the top, middle and bottom position of fuel element plate (FEP) were put into an anion exchange column containing 1 g of Dowex-Cl resin. The results of the separation process were 235U isotope identified according to their position in the irradiated U3Si2/Al fuel element plate i.e. top, middle, and bottom. It was found that the recovery values were 13265.7866 μg/gFEP; 6923.9248 μg/gFEP; and 12332.4175 μg/gFEP representing their position of top, middle and bottom respectively. It is concluded that the use of 9M of HCl solvent and 1500 ppm of Na2SO4 by anion exchange column method can improve uranium separation recovery from 68.2137 % to 95.7124 %. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7894 10.17146/urania.2017.23.3.3679 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 195-204 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7894/6028 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7895 2026-04-29T04:21:16Z urania:ART

PENGARUH VISKOSITAS SOL DAN PRESOLIDIFIKASI PADA PROSES GELASI EKSTERNAL DALAM PEMBUATAN GEL AMMONIUM DIURANAT MENGGUNAKAN SURROGATE CERIUM Sukarsono Rachmawati, Meniek Sol-gel process High temperature reactor fuel Cerium Zirconium External gelation Presolidification The external gelation process is used in the preparation of ammonium diuranat gel with processed into UO2 coated particles. UO2 coated particles are the core of high temperature reactor (HTR) fuels. This paper reported the research of gel preparation though external gelation process using a surrogate of zirconium stabilized cerium. The purpose of this study was to determine the effect of sol viscosity and presolidification related to gel roundness which was produced in the operation of the gelation unit in the laboratory of High Temperature Nuclear Reactor Fuel. The parameters studied were sol viscosity by varying the addition of additive and gelation process with presolidification (with NH3 gas stream) and without presolidification (without NH3 gas flow). The Ce-Zr gel can be prepared through a similar process of ADU gel preparation. A mixture of cerium and zirconium nitrate solution with Ce content of 12 mole %, was added by urea as a pH regulator and added by additive materials of tetra hydrofurfuryl alcohol (THFA) and poly vinyl alcohol (PVA). The solution was ready to be used in the gelation process. Sol solution was fed into the gelation column by pressing using compressed air. The sol solution entered the gelation column through the dripping node nozzle was vibrated with frequency of 90-110 Hz and certain amplitude. The droplets was passed on a column containing NH3 gas so that occurred presolidification on the surface of the droplets. The drops fall to the solution medium containing NH4OH in the gelation column. The compaction occurred due to the polymerization of cerium nitrate and zirconium nitrate at a pH of about 10. The gel was accommodated in a container located under the gelation column and allowed to be at least 2 hours to complete the solidification. The gel was washed successively with aqueous NH4OH, water and isopropyl alcohol and continued by drying. Operation of the gelation column to produce a round and uniform gel required adjustment the frequency, amplitude and flow rate of the sol solution. Operation of gelation columns at a frequency of 100 Hz using a sol with a viscosity of 97 cP and a flow rate of 25.8 mL/min produced a gel with smooth, uniformity and a relatively good sphere. For sol with a viscosity of 56 cP had relatively good flow rate was 28.8 mL/min and at a low viscosity of 46 cP, a flow rate of 30 ml/min is obtained. Operating the gelation device at frequencies of 110 Hz, 100 Hz and 90 Hz, produced a smooth, uniform and round gel at a sol flow rate of 27.9 mL/men, 25.8 mL/min and 21.2 mL/min. The gelation process with NH3 gas stream as a presolidative medium produced a more rounded and uniform gel than with no NH3 gas stream that resulted the non-round gel. From this studied it could be concluded that the viscosity of the sol and the vibration frequency of the dropout nozzle affects the flow rate of the sol but did not affect the gel roundness. The larger viscosity of solution, the smaller the flow rate of the sol to get a separate, uniform and round gel. The greater vibrational frequency also increases the flow rate to get a good gel. The gelation process without presolidification resulted a non-round gel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7895 10.17146/urania.2017.23.2.3229 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 69-86 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7895/6029 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7896 2026-04-29T04:21:16Z urania:ART

KARAKTERISTIK INGOT PADUAN U-Zr-Nb PASCA PROSES QUENCHING Masrukan Setiawan, Jan Microstructure U-Zr-Nb Quenching Phase Density The U-Zr-Nb heat-quenching experiment has been subjected to various Nb compositions. Ingots of U-Zr-Nb alloys with varying Nb compositions are 1%, 4% and 7% and a fixed 10% Zr composition is made by melting U, Zr and Nb metals in an electric arc furnace. The quenching experiment is intended to convert an unstable phase αU into stable γU phase. Quenching experiment is intended to convert an unstable phase αU phase into a stable γU phase. In quenching experiments, the U-10Zr-1Nb, U-10Zr-4Nb, U-10Zr-7Nb ingot ingredients were fed into box that made of steel plate and then fed into a heating furnace and heated at 900 C and held for 2 hours for then cooled rapidly in water. After quenched then subjected to phase testing and microstructure observation. The phase test is performed using X-ray diffraction (XRD) and the data is processed using HighScore software, while the microstructure observation using optical microscope. Test results with XRD showed that the U-10Zr-1Nb alloy before quenched consisted only of an αU phase, the U-10Zr-4Nb alloy consist of 23,1504% αU phases and 76,8495% γU phases, while U-10Zr-7Nb consist of 34,1873% αU phases and 65,8127% γU phases. For U-10Zr-1Nb alloys after quenched consist of an αU phase, the U-10Zr-4Nb alloy consist of 44% αU phases and 46% γU phases, the U-10Zr-7Nb alloy consisted of 19% αU and 81% phases γU. For U-10Zr-1Nb alloys after quenched consist of an αU phase, the U-10Zr-4Nb alloy consisted of 44,6711% αU phases and 55,3289% γU phase, the U-10Zr-7Nb alloy consist of 17,9918% αU phases and of 82,0082% γU phases. The results of the theoretical density analysis on the phases formed on U-10Zr-1Nb, U-10Zr-4Nb, U-10Zr-7Nb alloys both before and after quenched were obtained that ingot of U-10Zr-1Nb, U-10Zr-4Nb alloy U-10Zr-7Nb shows that the αU and γU phase densities decrease slightly if the Nb content is lower. Meanwhile, microstructure observation results showed that in the higher of Zr content (7% Nb) will form smaller grains (fine) after quenching. It can be concluded that the quenching process of U-10Zr-1Nb, U-10Zr-4Nb and U-10Zr-7Nb alloys will change the phase that is from the αU phase to γU and the change the microstructure formed. The microstructural changes show that in the higher Nb content, the grains that are formed become finer. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7896 10.17146/urania.2017.23.2.3522 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 87-96 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7896/6030 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7898 2026-04-29T04:21:16Z urania:ART

PENGARUH PELARUT ORGANIK PADA PROSES PERTUKARAN ANION DALAM PEMISAHAN URANIUM DARI LARUTAN PEB U3Si2/Al PASCA IRADIASI Anggraini, Dian Boybul, Boybul Yanlinastuti, Yanlinastuti Nugroho, Arif Kriswarini, Rosika Ginting, Aslina Br. U recovery anion exchange Methanol Dowex 1x8 U3Si2/Al post-irradiation Separation method of uranium from nuclear fuel solutions was developed to improve the quality of uranium-based analysis of post-irradiation nuclear fuel. The aim of this study was to determine the effect of organic solvent on the anion exchange process on the recovery of U from U3Si2/Al fuel solution post-irradiated. The organic solvent, in this case methanol, contained in solvent medium in the anion exchange column may be possible to increase the resin sorption of the uranium. The materials used are standard solution of uranil nitrat and supernatant solution of Cs separation process from the fuel solution. The uranium separation method from U3Si2/Al post irradiated fuel solution uses two columns of anion exchanger with two stages. The first column uses Dowex 1x8-Cl Dowex resin. Parameters varied in the first column are HNO3 concentrations ranging from 1M; 2M; 3M and 4M. The parameters variation in the second column is the ratio of HCl/methanol volume of 30:70 %; 20:80 % and 10:90 %. The result obtained a maximum U recovery of 90,68 % with feed solution composition in the first column containing 3 % HNO3 and methanol by 50 %volume, while in the second column with 6M HCl composition with methanol in 10 %:90 % ratio. In these conditions obtained uranium content in a solution of the middle PEB U3Si2/Al TMU 2.96 g/cm3 piece of 0.4725 μg. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7898 10.17146/urania.2017.23.2.3576 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 97-106 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7898/6031 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7899 2026-04-29T04:21:16Z urania:ART

ANALISIS KANDUNGAN CESIUM DAN URANIUM DALAM BAHAN BAKAR U3Si2/Al PASCA IRADIASI Boybul, Boybul Yanlinastuti, Yanlinastuti Nugroho, Arif Anggraini, Dian Kriswarini, Rosika Ginting, Aslina Br. 137Cs 235U U3Si2/Al fuel plate 2.96 gU/cm3 uranium density Separation of fission product isotopes of 137Cs and 235U has been done to determine content of 137Cs and 235U in irradiated U3Si2/Al fuel plate of 2.96 gU/cm3 density. The initial step in the determination of 137Cs and 235U content is sample cutting process. The cutting position was determined at the top middle and bottom part of the plate by an amount of 0.095 g; 0.1103 g and 0.1441g for each respective position. Each sample was dissolved in 25 mL of HNO3 6N dan HCl 6N to obtain irradiated U3Si2/Al solution. Separation was carried out by cation exchange and precipitation methods. The separation of 137Cs was done by pipetting 150 µL of the irradiated U3Si2/Al sample solution and adding it to 1000 mg of Lampung zeolite, while the separation by precipitation method was done by adding 700 mg of CsNO3 powder as carrier and HClO4 to the U3Si2/Al sample solution. In the cation exchange method, 137Cs would be bound by zeolite in the solid phase in the form of 137Cs-zeolite and uranium would stay in the liquid phase or supernatant. The precipitation was done in an ice bath at -4 oC to obtain 137CsClO4 precipitate. The effluent leaving the column was subjected to electrodeposition process for 235U content measurement with α-spectrometer. The 137Cs content obtained by cation exchange was 401.0335 µg/g FP (FP = irradiated fuel plate); 451.1094 µg/g FP and 343.9651 µg/g FP for each respective sample position by 99 % recovery, while the 137Cs content obtained by precipitation method was 393.4581 µg/g FP; 452.0525 µg/g FP and 330.7839 µg/g FP for each respective sample position by 98 % recovery. The 235U content was found to be 45208 µg/g FP; 50896 µg/g FP and 44336 µg/g FP for each respective sample position by 68 % recovery. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7899 10.17146/urania.2017.23.2.3568 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 107-116 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7899/6032 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7901 2026-04-29T04:21:16Z urania:ART

PENGARUH DENSITAS URANIUM TERHADAP UMUR BAHAN BAKAR NUKLIR DI DALAM REAKTOR RSG-GAS DITINJAU DARI ASPEK NEUTRONIK Octadamailah, Saga Supardjo Uranium density Fuel life Burn up Research reactor Neutronic Material Testing Reactor (MTR) is a research reactor with characteristics: small power, high thermal neutron flux and short life time fuel. Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) located in Serpong is an MTR type reactor initially operated with dispersion fuel of U3O8/Al with uranium enrichment of 19.75 % 235U and uranium density of 2.96 gU/cm3. In an effort to increase the lifetime of nuclear fuel, BATAN has done research on high density uranium fuel starting from U3Si2/Al fule and continued with UMo based fuel in line with investigations performed by worldwide researchers in the field of research reactors. Fuel of U3Si2/Al with a density of 2.96 gU/cm3 has been successfully manufactured and utilized as fuel for RSG-GAS to substitute the U3O8/Al fuel. Post irradiation examination has also been done for U3Si2/Al with a density of 4.8 gU/cm3 up to a burn up of 60 %, while research on UMo with 7 gU/cm3 density has been progressing into mini plate fabrication. Research on high density fuel has been focused on fabrication process, while little has been done on lifetime calculation or utilization period. Therefore, this research has performed a calculation on nuclear fuel lifetime using a paired program of ORIGEN and MCNP. The calculation results show that an increase in uranium density will extend the lifetime of fuel exponentially. The longest lifetime is achieved by UMo type, which extends to 651 days, and this relates to a lifetime extension of 219 days if the RSG-GAS is operated with UMo fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7901 10.17146/urania.2017.23.2.3550 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 117-126 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7901/6033 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7902 2026-04-29T04:21:16Z urania:ART

ANALISIS KRITIKALITAS BAHAN BAKAR NUKLIR BEKAS REAKTOR RSG-GAS PADA RAK BERBAHAN ALUMINIUM Artiani, Pungky Ayu Mirawaty, Mirawaty Heriyanto, Kuat Sub-criticality Spent fuel storage aluminum rack The use of stainless steel in nuclear spent fuels storage racks of interim storage facility potentially causes galvanic corrosion so material replacement of spent fuels storage racks needs to be considered. Galvanic corrosion is due to interaction between aluminum as in nuclear fuel of Multipurpose Research Reactor - G. A. Siwabessy (RSG-GAS reactor) and stainless steel as storage rack material for nuclear spent fuel. The purpose of this study is to assess the criticality of nuclear spent fuel storage racks using aluminum material to reduce galvanic corrosion effect. criticality guarantee is necessary to maintain the safety of nuclear spent fuel interim storage facility. Optimum criticality of aluminum storage rack is assessed by varying pitch size and calculating neutron absorption rate performed under normal condition (not an accident). Criticality (keff) calculation is performed using Monte Carlo N-Particle version 6 (MCNP6) code. The model used in this study is a 3-dimensional model where racks are filled fully with spent fuels in the storage pool. The calculations show that nuclear spent fuel storage rack stainless steel materials (keff = 0.6810) compared to aluminum (keff = 0.7709) in the size of 127 mm pitch, keff increased by 13.20%, but still within the value of safety requirement (keff ≤ 0.95). keff is affected by pitch size, where the reduction in the size of 1 mm pitch can increase the keffof 14.24%. Neutron absorption rate also affects keff, where neutron absorption rate in aluminum storage racks smaller than stainless steel. Storage racks using aluminum materials meet safety aspect to use as a storage rack in nuclear spent fuel interim storage facility because it has a keff less than 0.95 on a pitch size more than 112 mm (keff = 0.9196). Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7902 10.17146/urania.2017.23.2.3280 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 127-138 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7902/6034 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7903 2024-12-23T01:46:24Z urania:ART

DIP-COATING PROCESS OF ZIRCALOY-2 FUEL CLADDING WITH COLLOIDAL GRAPHITE Mutiara, Etty Rachmawati, Meniek Zircaloy Cladding Dip-coating Colloidal graphite ZrC The intensive research on high discharge burn-up of Light Water Reactor (LWR) fuel element have been continuously performed due to the extension of fuel element’s utility life. One of this research was allowing for alteration of the existing zirconium-based clad system through coating. A coating technique with the coating layer thickness of 10 - 30 µm will improve the corrosion resistance of cladding without changing the dimension of cladding. The scope of this current research is to obtain the zircaloy-2 cladding coated with ZrC layer by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens undergo heating process at 700 oC, 900 oC and 1100 oC respectively in Argon gas atmosphere for 1 hour and are subsequently characterized by optical microscope and XRD. The optical microscope images show that the coating layers thickness are increased as the heating temperature increased. The coating layers thickness are 10 μm, 20 - 40 μm and 100 μm for the specimens heated at 700 oC, 900 oC and 1100 oC respectively. The calculated diffusivity of carbon into zircaloy-2 cladding for the coated specimens at 700 oC, 900 oC and 1100 oC are 3.10216E-11 cm2s-1; 3.60479E-11 cm2s-1 and 4.00613E-11 cm2s-1 respectively. From XRD examination analysis reveals that the ZrC phase appears in the specimens heated at 1100 oC but it is not the case for specimens heated at both 700 oC and 900 oC. The coating layer of specimens heated at both 700 oC and 900 oC mostly consists of carbon. At these heating temperatures, carbon atoms have diffused into zircaloy-2 and substituted the zirconium atoms with a limited occupation to form C-Zr solid solution. At the temperature of 1100 oC, due to the increase in vibration energy, the carbon atoms have enough energy to diffuse to form the carbide phase. Heating process at higher than 700 oC, however, will degrade the zircaloy-2 cladding. It is concluded that the dip-coating process of zircaloy-2 cladding in graphite colloid with subsequent high temperature heating is not the proper method to gain the ZrC-coated zircaloy-2 cladding. Therefore, as for the future works of this research, the others method should be searched and investigated to obtain the proper ZrC coating process on LWR zircaloy cladding which fulfills the dimension and quality requirements. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7903 10.17146/urania.2017.23.1.3148 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7903/6036 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7904 2024-12-23T01:39:49Z urania:ART

PERHITUNGAN BURN UP PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI POTONGAN BAGIAN MIDDLE Ginting, Aslina Br. Supardjo, Supardjo Yanlinastuti, Yanlinastuti Indaryati, Sutri Boybul , Boybul U3Si2/Al density 4,8 gU/cm 3 cesium uranium separation burn up BURN UP CALCULATION OF THE MIDDLE PART OF IRRADIATED U3Si2/Al FUEL PLATE WITH 4.8 gU/cm3 DENSITY. Post-irradiation examination with non-destructive and destructive methods is required to study the performance of U3Si2/Al fuel during irradiation at the G. A. Siwabessy Multipurpose Reactor (RSG-GAS). The purpose of the burn-up calculation is to discover the 235U content that burned in a fuel plate of U3Si2/Al during use in RSG-GAS and to prove the burn-up obtained using destructive method by calculation using Origen Code. Therefore, this research performed a burn-up calculation of U3Si2/Al fuel plate (with a density of 4.8 gU/cm3) using destructive method through physicochemical analysis. The physicochemical analysis was aimed to determine the isotope content of 235U, 134Cs, and 137Cs in the irradiated U3Si2/Al fuel plate. The fuel plate was cut at the middle using a diamond cutting machine in a hot cell to obtain a sample cut with a weight of M-1 = 0.056 g, and a duplicate cut was also prepared with a weight of M-2 = 0.075 g. The sample was then dissolved using HCl 6 N and HNO3 6N. The results of dissolution was UO2(NO3)2 solution containing isotopic fission such as cesium and uranium. The solution of UO2(NO3)2 was pipetted at 1 mL and then transferred from the hot cell to R.135 (outside the hot cell) to separate 134Cs and 137Cs from 235U by a cation exchange method using zeolite from Lampung. The solution of UO2(NO3)2 was pipetted at 100 µL into a vial in duplicates and then 1000 mg zeolite was added. The cation exchange process was then carried out by stirring for 1 h. The separation process resulted in 134Cs and 137Cs isotopes in the solid phase, while uranium isotopes (238U, 235U, 234U), plutonium (239Pu, 238Pu) and other isotopes were in the liquid phase. Separation of uranium from other isotopes was done by anion exchange column method using 1200 mg of Dowex resin. The isotope content of 235U, 134Cs, 137Cs, 239Pu, and 238Pu was then measured using α/- spectrometer, the values of which were subsequently used for burn-up calculation. The result of the burn up calculation is 41.6565 ± 1.687%. The value of burn-up is not significantly different from the burn-up obtained from the calculation using Origen Code, which is 40%. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7904 10.17146/urania.2020.26.2.5864 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 91-106 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7904/6038 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7905 2024-12-23T01:46:24Z urania:ART

EFFECT OF CURRENT, TIME, TYPE OF FEED AND CATHODE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION Sigit Widodo, Ghoib Wasito, Bangun Basuki, Kris Tri Fahrunissa Electroplating Uranyl nitrate Effluent process Pt/Pt electrode Pt/SS electrode Electroplating of uranyl nitrate and process effluent has been carried out in order to collect uranium contained therein using Pt/Pt and Pt/SS electrodes at various current and time values. The materials used as electrode were Pt (platinum) and SS (Stainless Steel). Feed solution of 250 mL was entered into a beaker glass equipped with Pt anode - Pt cathode or Pt anode - SS cathode and connected to direct current source from DC power supply at various current values and for various time periods so that uranium precipitated and deposited onto the cathode. Upon deposition completion, the cathode was removed and weighed to determine weight of precipitates, while the solution was analyzed to determine uranium concentration decrease after electroplating process. The experiment shows that a relatively good time to acquire uranium deposit at the cathode was 1 hour at 7 amperes, with uranyl nitrate as feed and Pt (platinum) as cathode. In these conditions, uranium deposit attached to the cathode amounted to 74.96% of the original weight of uranium oxide in the feed or 206.5 mg by weight. The use of Pt cathode for uranyl nitrate feed, and SS and Pt cathodes for effluent feed resulted in uranium specific weight at the cathode of 12.99 mg/cm2, 2.4 mg/cm2 and 5.37 mg/cm2 respectively for 7 Ampere current for 1 hour electroplating time. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7905 10.17146/urania.2017.23.1.3155 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-22 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7905/6037 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7906 2024-12-23T01:46:24Z urania:ART

PROSES EKSTRAKSI-STRIPPING UO2(NO3)2 BERIMPURITAS HASIL PELARUTAN DARI YELLOW CAKE Muchsin, Anwar Widodo, Ghoib Extraction Stripping Feed rate Ratio of TBP/kerosene Acidity Uranyl nitrate/UO2(NO3)2 is a basic material/bait that is used as fuel both a research reactor fuel and power reactors. The process of extraction-stripping necessary to UO2 (NO3)2, because in the UO2(NO3)2 is still contained impurities are entrained during the dissolution process yellow cake. To obtain a solution of UO2 (NO3)2 which meets the requirements of pure nuclear (nuclear grade), it should be a process of purification by extraction-stripping using three (3) parameters of the feed flow rate, the acidity of the feed to the extraction process, and acidity in the stripping process. The mechanism of the extraction process takes UO2 (NO3)2 will go into the organic phase TBP, whereas impurity are in the water phase (refined). Uranium in the organic phase stripping process is carried out by determining the acidity (as a parameter), in order to obtain a mixed solution UO2 (NO3)2 pure nuclear (nuclear grade). As a bait used uranyl nitrate solution with levels of uranium amounted to 225.6282 gU/L. The results extraction process obtained the rate of feed water 15 L/h, the acidity of the feed to the extraction process by 3 M and obtained levels of uranium amounted to 48.5365 gU/L, while the stripping process is obtained UO2 (NO3)2 amounted to 64.7860 gU/L the acidity of 0.04 M. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7906 10.17146/urania.2017.23.1.3172 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 23-32 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7906/6039 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7907 2024-12-23T01:46:24Z urania:ART

ANALISIS KUAT SUMBER NEUTRON DAN PERHITUNGAN LAJU DOSIS NEUTRON TERAS AWAL RDE Suwoto Adrial, Hery Zuhair TRISO Pebble RDE Neutron dose rate ICRP NCRP The reactor core RDE (Experimental Power Reactor) has non-annular cylindrical shape with coated fuel particles TRISO kernel in the spherical form (pebble) and cooled by helium gas. The RDE reactor core design is very safe, because it adopts the high-temperature reactor technology (HTGR) with inherent passive safety. The designed input helium temperature of RDE core is about 250 °C and the output of helium gas temperature helium of the reactor core of 700 °C. Beside of generating electricity, the RDE reactor is designed to generate high temperature gas that can be used for any purposes such as cogeneration (heat process application research). The spherical fuel called as pebble containing thousand of kernel TRISO-coated fuel particles of uranium dioxide 17 % (235U) enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner part pyrolytic carbon layer (IPyC, Inner Pyrolitic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer part (OPyC, Outer Pyrolitic Carbon). Analysis of neutron source strength and preliminary calculation of neutron dose rate on the RDE core were performed using Monte Carlo MCNP5v1.2 code. Double heterogeneity treatment in TRISO-coated fuel particles kernel and the pebble of the core are applied. Utilization of EGS99304 code to determine 640 amount of energy group structures (SAND-II neutron group structures) are used in the neutron spectrum calculation in the reactor. The reactor core is divided into 100 zones (10 in radial and 10 in axial directions). An analyses calculation result shows that the neutron source strength of the core is about 8.47027X1017neutron/sec. The neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection (ICRP) and National Council on Radiation Protection and Measurements (NCRP). The preliminary calculations result of neutrons dose rate using ICRP-21 and NCRP-38 conversion factor for radiation workers in the radial direction in the shield biological are already weakened as low as 6.69915 and 6.9964 μSv/h, respectively, at a position of 215 cm from the center part of the core. This result analysis indicate that radiation workers are safe and protected from neutron radiation sources in accordance with the regulatory requirements. Overall, from the calculation and analysis, it appears that the radiation and biological shielding model meets to the standards of radiation safety required. The results of this study can be used to complete the design data thickness of shielding of the reactor. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7907 10.17146/urania.2017.23.1.3119 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 33-44 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7907/6040 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7908 2024-12-23T01:46:24Z urania:ART

ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE Udiyani, Pande Made Kuntjoro, Sri Routine discharge Radionuclide Source terms RDE HTGR Experimental power reactor (RDE), which is planned to be constructed by BATAN, is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products to be remained in the core. To ensure the safety for workers and public surroundings of the RDE site and to meet the regulatory body requirements for construction and operation, research is needed for environmental radiation in normal and abnormal condition. This study is focused on normal operating condition, while for the accident condition study will be carried out after the final design has been completed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source terms released into the environment based on postulation of normal or routine operation of the RDE-10 MWth. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitations during the fabrication. Mechanisms of fission products release from the fuel to the environment were created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated by the use of ORIGEN-2 whose library has been modified for HTGR type, while assumptions on TRISO fuel defects and release fraction for each compartment of the RDE safety systems are defined by a reference parameter. The results show that the important source terms of RDE are group of noble gases (Kr and Xe), halogen (I), Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTR-10 source terms with the same power. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7908 10.17146/urania.2017.23.1.3160 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 45-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7908/6041 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7909 2024-12-23T01:46:24Z urania:ART

PENENTUAN KEMURNIAN RADIOKIMIA 99mTc-MIBI SECARA CEPAT DAN PRAKTIS MENGGUNAKAN DOSE CALIBRATOR Maskur, Maskur Lestari, Enny Sarmini, Endang Tahyan, Yayan Putra, Amal Rezka Kurniasih, Dede Gunawan, Adang Hardi Radiochemical purity Radiopharmaceutical 99mTc-MIBI Gamma counter Dose calibrator Generally, the determination of radiochemical purity begins with Radiochemical separation using thin layer chromatography (TLC), and then measurement using a gamma counter, but this method cannot be applied in hospitals since there is only a dose calibrator available in the hospitals. Therefore, we conducted such an investigation using an alternative method which is more practical and faster, ie begins with radiochemical separation using extraction method, and then measurement using a dose calibrator. The determination was carried out in which 99mTc-MIBI was extracted using a mixture of saline and chloroform with volume ratio of 1:1, the mixture was then extracted. The 99mTc - MIBI was in organic phase (chloroform), whereas 99mTcO2 and 99mTcO4- were both in water phase (saline). Then, each phase of the sample was taken with the same volume, and then the radioactivity was measured using a gamma ionization chamber and the radiochemical purity was calculated. Radiochemical purity with extraction process method Compared measuring using standard methods of TLC Furthermore, the results of the extraction were compared with the standard method of TDC. The results showed that the radiochemical purity of 99mTc-MIBI using the two methods were almost the same, ie =98.34 % ± 0, 65 % for TLC standard method and = 97.15 % ± 0,56 % for extraction method with good measurement precision. The time of determination for extraction method was 20 minutes and 180 minutes for TLC standard method respectively. These results indicated that the extraction method can be used as an alternative method for determination of radiochemical purity of 99mTc-MIBI radiopharmaceutical. The advantage of extraction method that it is faster compared to the TLC Method. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7909 10.17146/urania.2017.23.1.3195 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-68 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7909/6043 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7910 2024-12-23T01:39:49Z urania:ART

PENENTUAN BERAT ISOTOP CESIUM DAN URANIUM DALAM PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 Yanlinastut, Yanlinastut Indaryati, Sutri Boybul, Boybul Supardjo, Supardjo Ginting, Aslina Br. fuel element plate of U3Si2/Al separation cation-anion cesium uranium DETERMINATION OF WEIGHT OF CESIUM AND URANIUM ISOTOPES IN IRRADIATED U3Si2/Al FUEL PLATE WITH 4.8 gU/cm3 DENSITY AND 40% BURN UP. The fuel plate of U3Si2/Al with a density of 4.8 gU/cm3 was irradiated in the RSG-GAS multipurpose reactor with 40% burn- up. During irradiation, the amount of isotopes produced by fission reaction, such as cesium, uranium and other isotopes can be used for burn-up calculation to determine the fuel performance. Physicochemical analysis was performed to determine the weight of cesium and uranium isotopes in the irradiated fuel. The irradiated plate of U3Si2/Al was cut at the middle section (duplo) with a weight of M-1 = 0.065 g and M-2 = 0.075 g and dissolved in HCl 6N and HNO3 6N. The resulted was pipetted at 100 µL and then added into each of 2 mL solvent of HNO3, H2SO4 and HCl 0.1N. The separation of cesium was then carried out with the addition 1000 mg of zeolite from Lampung. Cesium isotope obtained from the separation was bound with zeolite in the solid phase while uranium was in the liquid phase. Cesium isotopes in the solid phase and in the liquid phase were measured by a spectrometer- for their radioactivity for 10000 seconds. The measurement was indicated by spectra of 134Cs at 604.7 and 795.8keV energy and 137Cs at 661.7keV. Meanwhile, the separation of uranium in the liquid phase was carried out using anion exchange column method with the addition of Dowex resin. The results of the separation were indicated by some spectra of U isotopes at various energies, 238U at 4,1943 MeV, 235U at 4,397 MeV, 236U at 4,494 MeV and 234U at 4,777 MeV. Weight calculation from the measurement data shows that the weight of 137Cs in 0.056 g of the irradiated U3Si2/Al fuel samples in HNO3, H2SO4 and HCl solvent are 4.08E-05 g; 4.53E-05 g ,and 4.53E-05 g respectively, while 134Cs weight are 4,84E-08 g; 5,30E-08 g and 5,20E-08 g respectively. Meanwhile, 137Cs isotope weight in 0.075 g fuel plate of U3Si2/Al are obtained by 5.93E-07 g; 5.90-05 g, and 5.88E-05 g, while the isotope weight of 134Cs in HNO3, H2SO4 and HCl solvent are 5,23E-07 g; 5,29E-07g and 5,008E-07g respectively. The isotopes weight of 238U, 235U, 236U and 234U in 0.056 g of fuel plale are 6.45E-03 g; 2.57E-03 g; 2.42E-06 g; and 1.90E-05 g, and the weight in 0.075 g fuel plale are 7.41E-03 g; 2.40E-03 g; 2.91E-06 g; and 2.17E-05 g. The isotopes weight of 134Cs, 137Cs, 238U, 235U, 236U and 234U in 0.056 g and 0.075 irradiated U3Si2/Al fuel plate obtained were relatively similarly. Data of isotopes weight are then used as input data for burn up calculation. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7910 10.17146/urania.2020.26.2.5874 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 107-120 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7910/6055 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7911 2024-12-23T01:37:56Z urania:ART

KOMPOSISI, STRUKTUR DAN DENSITAS PADUAN U-7Mo-xSi Supardjo Kadarjono, Agoeng Setiawan, Jan Suryaman, Ganisa Kurniati U-7Mo-xSi alloy High density nuclear fuel Composition Structure Density The structure of U-7Mo-xSi (x = 1, 2, and 3%) alloy was studied to obtain a new uranium alloy suitable for use as a candidate for high-density uranium research reactor fuel. The U-7Mo-xSi alloy was made by melting technique using an arc furnace. The characterization of the alloy includes chemical composition, density, and structural phases to identify the phases contained in the alloy. The hardness test was also done. The density characterization results show that U content and density of the alloy decrease with increasing of Si content in the U-7Mo-xSi alloy. On the other hand, the hardness of the alloy increases with increasing of Si content. The X-ray diffraction pattern of U-7Mo-xSi alloy was observed from diffraction angle of 25° to 95°. The results indicate the presence of U-γ, U3Si2, and U3Si2Mo phases. The higher the Si content, the more U3Si2 phase is formed and the more the hardness increases due to the formation of bonds between Si and Si. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7911 10.17146/urania.2021.27.2.6386 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-66 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7911/6044 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7912 2024-12-23T01:37:56Z urania:ART

THE EFFECT OF ANNEALING AND COLD FORGING ON MICROSTRUCTURE AND HARDNESS PROPERTIES OF Al-SiC COMPOSITE : A PRELIMINARY STUDY Adityawardhana, Yudhistira Zulfia, Anne Adjiantoro, Bintang Hasbi, Muhammad Yunan Al-SiC composite Annealing temperature Cold forging Hardness Microstructure Aluminum Metal Matrix Composites (AMMCs) are among the exciting materials that have an extensive function in various applications. Reinforcement in the fabrication process of Al composites can produce superior properties such as high strength, good fracture resistance, and of course, lightweight. Therefore, many studies are interested in revealing the characteristics of Al composite materials through various methods and variations of reinforcement. This research was a preliminary study with a scope of work that includes observation on the effects of annealing and cold forging processes on microstructure morphology and hardness properties of SiC nano-ceramic reinforced of Al composites. The aluminum used in this study was a 7xxx series aluminum alloy. The fabrication process was carried out by stir-squeeze casting method. Microstructure analysis was conducted by optical microscopy and Scanning Electron Microscopy (SEM) equipped with Emission Dispersion Spectroscopy (EDS). The hardness properties of the Al-SiC composite were examined by micro-Vickers hardness testing. This paper reports that annealing process influences grain refinement and hardness properties of the Al-SiC composite. The sample experiencing cold forging shows improvement in the hardness value. Increase of hardness value by forging after annealing may be promoted due to grain compression as dislocation mechanism effect. Comprehensive research is required to discover other potentials of Al-SiC composite materials. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7912 10.17146/urania.2021.27.2.6396 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 67-74 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7912/6045 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7913 2024-12-23T01:37:56Z urania:ART

PENGARUH JENIS PREKURSOR DAN METODE PRENETRALISASI PADA MORFOLOGI PERMUKAAN KERNEL TERSINTER ZIRKONIA Sarjono Rahmadi, Gagad Mutiara, Etty Mujinem Haryadi, Dedi Kernel Zirconia Precursor Preneutralization Surface roughness Research activities have been carried out on the effect of precursor types and preneutralization methods on the surface morphology of sintered zirconia kernels. Surface morphology with minimum roughness is desired in the development of TRISO coating system. Several batches of broth solution were prepared by several routes using different types of precursors and preneutralization methods. The types of precursors used were zirconium oxynitrate and zirconium nitrate pentahydrate and the preparation of sol was carried out by precipitation-redispersion and urea decomposition. The experimental stages included sol preparation, broth preparation, gel-casting process, aging-washing-drying process, de-watering process, heat treatment of calcination and sintering, and visual examination of surface morphology. Sphericity measurements were carried out on dry spherical gels to study the level of saturation of the sol and its ability to convert instantly upon reaction with NH3 gas in the gel-casting process. The results of SEM visual observations of the surface of the sintered zirconia kernel show that the lowest surface roughness was given by the kernel from zirconium nitrate pentahydrate precursor with urea decomposition pre-neutralization method, and the roughest surface was given by the kernel from zirconium oxynitrate precursor with pre-neutralization method of precipitation-redispersion. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7913 10.17146/urania.2021.27.2.6406 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 75-84 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7913/6046 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7914 2024-12-23T01:37:56Z urania:ART

KARAKTERISASI TERMAL PEB U-7Mo/Al SEBAGAI BAHAN BAKAR REAKTOR RISET Nisa, Sayyidatun Saputri, Ariyanti Indaryati, Sutri Noviarty, Noviarty Supardjo Ginting, Aslina Br. Thermal properties DSC UMo/Al fuel element plate Research reactor UMo/Al-based fuel is a candidate for research reactor fuel and as an alternative to U3Si2/Al fuel. This is because the UMo/Al alloy has a higher uranium density of 16.4 g/cm3 compared to U3Si2/Al of 12.2 g/cm3. In addition, the UMo alloy has a low neutron absorption cross section and its reprocessing process is relatively easy. The U-7Mo/Al alloy is able to maintain the γ-U phase during irradiation process so that the stability of the fuel in the reactor is maintained. The U-7Mo alloy has been fabricated into a U-7Mo/Al mini fuel element plate. The purpose of this study was to determine the thermal characteristics of U-7Mo/Al as inputs to the research reactor fuel fabricator in the manufacture of UMo/Al-based fuel. Thermal characterization was carried out for uranium, molybdenum, UMo and the fuel element plate of U-7Mo/Al using Differential Scanning Calorimetry (DSC). The thermal characterization results show that the U-7Mo/Al fuel element plate is thermally stable up to 600 °C because it undergoes an endothermic thermochemical reaction at a temperature of 632.962 °C with DH = 144.318 J/g. At 656.186 °C, the Al matrix and AlMg2 cladding are fused in the U-7Mo/Al with DH = 144.318 J/g. At temperatures of 952.810 °C to 1106.050 °C, the molten Al matrix diffuses with UMo to form a metastable U(Mo,Al)X compound and heating to a temperature of 1211.760 °C forms a UAlX layer (UAl2, UAl3 and UAl4) with DH = –22,721 J/g. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7914 10.17146/urania.2021.27.2.6420 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 85-92 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7914/6047 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7915 2024-12-23T01:37:56Z urania:ART

STUDI KELAYAKAN YELLOWCAKE HASIL SAMPING PETROKIMIA GRESIK SEBAGAI BAHAN BAKU PROSES PEMURNIAN DAN KONVERSI DI FASILITAS PILOT CONVERSION PLANT Saputra, Ade Yusnitha, Erilia Abdurrosyid, Imam Langenati, Ratih PCP Yellowcake purification Yellowcake conversion Petrokimia's yellowcake The Pilot Conversion Plant (PCP) facility has been successfully used for the purification and conversion of commercial yellowcake from Cogema – France. Thus, to support self-reliance of the nation, an assessment of the purification and conversion process of domestic yellowcake was carried out. One of domestic sources of yellowcake is the by-product of a process carried out at Petrokimia, Gresik. This study was carried out by comparing the characteristics of the chemical and physical properties of Petrokimia's yellowcake to those of Cogema's yellowcake which has been used at the PCP. Primary information was obtained from analysis results while secondary information was obtained from related literatures. The information obtained is also used to determine the process parameters needed by the PCP for the purification and conversion process. The results of this study show that Petrokimia's yellowcake is suitable to be used as the feed of purification and conversion process at the PCP facility. In addition, Petrokimia's yellowcake is easier to handle although requires more nitric acid than Cogema yellowcake dissolution. The parameters values of Petrokimia's yellowcake purification and conversion process that have been obtained include dissolution of yellowcake, purification of uranyl nitrate, uranyl nitrate concentrating, precipitation of uranyl nitrate into ammonium diuranate (ADU), drying of ADU powder, calcination of ADU into U3O8 powder and reduction of U3O8 powder into UO2 powder, which are no different from the parameters values for Cogema's yellowcake processing. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7915 10.17146/urania.2021.27.2.6424 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 93-102 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7915/6048 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7916 2024-12-23T01:37:56Z urania:ART

EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Heriyanto, Kuat Sudjadi, Usman Rachmadetin, Jaka Purwanto, Yuli Artiani, Pungky Ayu Chotijah, Siti Neutron shielding Cadmium Stainless steel Spent fuel Development of dry casks is necessary to support national strategy for the management of spent fuels. One of the requirements for the dry cask is shielding performance for neutrons emitted by the spent fuels. The objectives of this study were to determine the emitted neutrons by the spent fuels generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used for dry cask for the spent fuels.  The determination of the emitted neutrons was carried out using Origen 2.1 software while the evaluation of the neutron shielding performance was done using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor are 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. It can be concluded that evaluation of neutron shielding performance of Cd-SS316L can be done using Origen 2.1 software for the neutron emission determination and MNCP 5 software for the neutron shielding performance by simulation. This study shows that the Cd-SS 316L alloy is worth further study to develop the dry cask design for the GAS research reactor Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7916 10.17146/urania.2021.27.2.6426 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 103-110 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7916/6049 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7917 2024-12-23T01:38:34Z urania:ART

PENGARUH DISTRIBUSI UKURAN SERBUK UO2 PENGAYAAN 2%–5% PADA DENSITAS PELET UO2 Mutiara, Etty Hutapea, Odi Buana Dewayatna, Winter Yulianto, Tri PWR UO2 pellets Powder size distribution Density Compaction Sintering A study of UO2 pellet fabrication using 2%, 3%, 4% and 5% enriched UO2 powder without conditioning has been carried out. There is no published research on the relationship between UO2 powder size distribution and UO2 pellet density in the range of powder sizes studied in this work. Therefore, through this research the influence of size distribution and weight fraction of UO2 powder on various densities was studied to obtain UO2 sintered pellet with a density that meets the requirements as PWR fuel. UO2 pellets were fabricated through cold compaction with variation in pressure of 2, 3 and 4 ton/cm2 and sintering process at 1700 °C for 3 hours. Each UO2 feed powder with its size distribution and weight fraction was directly fabricated into pellets without conditioning.The powder sieving process shows that the 2% and 4% enriched UO2 powders have a monomodal powder size distribution while the 3% and 5% enriched UO2 powders are trimodal size. The green pellets density of 2%, 3%, 4% and 5% enriched UO2 powders are 54.3%TD, 53.19%TD, 52.45%TD and 52.33%TD respectively. The density of sintered pellets from 2% and 5% enriched UO2 powders are 91.13-91.70%TD and 75.33-77.57%TD. The sintered pellets of 3% and 4% enriched UO2 powders were cracked or broken so that the density could not be measured. Based on the pellet density data, it is concluded that the 2% enriched UO2 powders with monomodal size distribution that are dominated by fine powder smaller than 75µm has better compaction capability and sintering ability than other UO2 powders. This research needs to be continued to determine the proper size distribution of the UO2 powder and its weight fraction as compaction process feed in order to obtain higher sintered pellets density that meet the requirements of PWR fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7917 10.17146/urania.2021.27.1.6333 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-20 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7917/6050 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7918 2024-12-23T01:38:34Z urania:ART

PENGARUH MEDIA PENDINGIN TERHADAP KARAKTERISTIK MEKANIK DAN STRUKTUR MIKRO PADA PELAT BAJA KARBON RENDAH Wicaksono, Iskandarsyah Ismarwanti, Sri Setiawan, Jan Susetyo, Ferry Budhi Syamsuir Heat treatment Cooling media Mechanical properties Low carbon steel plate Low carbon steel has high ductility and is easy to shape, so it is widely used in industry and construction. However, this material has low mechanical strength and hardness, so it is easy to wear. Mechanical properties can be improved by heat treatment, namely heating at a specific temperature and holding it for some time and then cooling. Cooling media will affect the level of enhancement in its mechanical properties. Therefore, in this study, observations were done on the mechanical characteristics and microstructure of low carbon steel after heat treatment and cooling on various cooling media. The heating process was carried out at a temperature of 950 °C with a holding time of 15 minutes. The sample is then cooled with a variety of coolants (aqua dest, cooking oil, oil and open-door furnace). The results of the observations on samples of low carbon steel plates (0.0524%) shows that the sample without heat treatment has the highest stress value of 375.75 MPa and the open furnace door cooling has the lowest stress value of 200.53 MPa. The highest strain value occurred in the open-door cooling of 38.77% and the lowest strain value of 25.49% occurred in the cooling with cooking oil. Observation on the material's microstructure before and after heat treatment shows ferrite and pearlite. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7918 10.17146/urania.2021.27.1.6187 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 21-28 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7918/6051 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7919 2024-12-23T01:38:34Z urania:ART

KARAKTERISASI KANDUNGAN URANIUM DAN UNSUR JEJAK PELET SINTER UO2 UNTUK FORENSIK NUKLIR Torowati Ngatijo Rahmiati Mustika, Deni Yusnitha, Erilia Yulianto, Tri Setiawan, Jan Uranium Sintered pellets Fingerprints Nuclear forensic Nuclear forensics is one of the important tools in nuclear security related to law enforcement. Nuclear materials and radioactive materials have potential hazards in terms of safety, health, up to threats to the nuclear security. In nuclear forensics, fingerprints are the characteristics of nuclear materials and radioactive materials which provide information on the origins of the nuclear materials, from where data of nuclear materials and radioactive materials are expected to be collected. The data can be obtained from characterization by physical and chemical testing. The physical tests include visual observations, dimensions, and phases while chemical tests cover elemental analysis of the nuclear materials and analysis of elements concentration in the nuclear materials. In this study, analysis of uranium and trace elements contents in uranium dioxide pellets was carried out in PTBBN - BATAN with the purpose to collect the fingerprint for nuclear forensics. The analysis of uranium content was conducted by potentiometric titration and the trace elements content was analyzed by atomic absorption spectrophotometric method. The results show that uranium content of the nuclear materials tested has the average results from 87% to 88%, indicating that the materials are UO2. The analysis of trace elements content was done to UO2 of varied enrichment level and the results show that the trace elements are not similar, which informs about the level of enrichment of the uranium dioxide pellets. These data can be used as fingerprint for nuclear forensics that can assist the investigators in the identification of nuclear forensics when misuse of nuclear materials occurs. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7919 10.17146/urania.2021.27.1.6224 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 29-36 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7919/6052 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7920 2024-12-23T01:38:34Z urania:ART

PEMBUATAN SENYAWA BERTANDA 131I-MIBG DENGAN PENINGKATAN DOSIS TERAPI Maskur, Maskur Chairuman Sugiharto, Yono Asyikin, Khoirunnisa Fauziah Triningsih Sulaiman Labeled compounds of 131I-MIBG Radioactivity concentration Therapeutic doses Nowadays, the labeled compound of 131I-MIBG has been successfully produced in Indonesia. However, the concentration of radioactivity and dose of radioactivity per vial is still small when compared to commercial products from abroad available in the market. Currently, the dose of radioactivity achieved is 25 mCi/vial with a concentration of 5 mCi/mL. Problems arise if the radioactivity and dose of radioactivity per bottle are enlarged, where the product becomes unstable and breaks down because of the effects of radiolysis. The generally global commercial products contain a dose of radioactivity = 100 mCi/vial with a concentration of 10 mCi/mL. By this research efforts have been made to increase the concentration and dose of radioactivity per bottle. The method used was by increasing the amount of benzyl alcohol from 9 μL/mL to 10 μL/mL of 131I-MIBG solution. Other than that, the addition of phosphate buffer which was originally added before the purification process was switched to after the purification process. The result of labeling was 131I-MIBG 153 - 254 mCi with process yield between 33.60 and 51.94%. The radioactivity concentration was quite large, i.e., between 9.80 to 25.40 mCi/mL with radioactivity doses ranging from 100 to 222 mCi/vial. However, to adjust to market needs, the I-MIBG of Batan product was then packaged in 100 mCi/vial with a concentration of 5-13 mCi/mL.This workaround shows that the synthesis of labeled compound of 131I-MIBG for scale up therapeutic dose has been successfully carried out. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7920 10.17146/urania.2021.27.1.6358 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 37-46 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7920/6053 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7921 2024-12-23T01:38:34Z urania:ART

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK Luthfi, Wahid Pinem, Surian Neutronic parameter Critical boron concentration Power peaking factor SRAC code system Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choose because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction and Power Peaking Factor (PPF), and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7921 10.17146/urania.2021.27.1.6238 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 47-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7921/6054 Copyright (c) 2021 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7922 2024-12-23T01:39:49Z urania:ART

PENENTUAN UMUR YELLOW CAKE SECARA RADIOKRONOMETRI Noerpitasari, Erlina Fatimah, Syamsul Haryati, Iis Yanlinastuti, Yanlinastuti Setiawan, Jan Boybul, Boybul Nugroho, Arif Yellow Cake thorium radiochronometry YELLOW CAKE AGE DETERMINATION BY RADIOCHRONOMETRY. Radiochronometric dating of yellow cake has been carried out. The aim of this study was to obtain data on the age of yellow cake for compiling a database for nuclear forensics library. The radiochronometer used was228Th/232Th. 232Th is thorium which naturally co-exists in yellow cake and will reach a decay equilibrium to 228Th. In the yellow cake production process, however, the equilibrium is disturbed so that the "time" will be zeroed again. The test samples were yellow cake from Cogema, from PTBGN mining in Kalan, and by-products from Petrokimia Gresik. The steps of processes include dissolving the sample, separating thorium from uranium by cation exchange chromatography using dowex 50W-X8 resin, electrodeposition of thorium and isotopes analysis using an alpha spectrometer. Age determination was performed by calculating the 228Th/232Th activity ratio and decay formula iteration. Based on the analysis of the 228Th and 232Th isotopes using an alpha spectrometer and the results of age calculations, the age of Cogema yellow cake was not able to be determined because it does not contain 232Th. The age of Petrokimia Gresik yellow cake is 9.90 years with a bias of 20.35 years and the age of PTBGN yellow cake is 12.85 years with a bias of more than 20.15 years when compared to the known estimated production time. The bias obtained is higher when compared to the results determined by previous researchers by the same method,i.e., a bias of less than 10 years. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7922 10.17146/urania.2020.26.2.6057 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 121-130 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7922/6056 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7926 2024-12-23T01:39:26Z urania:ART

PENGARUH ARAH PENGEROLAN DAN SHEAR CUTTING TERHADAP KARAKTER PADUAN AlMg2 SEBAGAI MATERIAL CLADDING BAHAN BAKAR REAKTOR RISET Sungkono, Sungkono Ismarwanti , Sri rolling direction n, shear cutting AlMg2 alloy tensile strength hardness microstructure THE INFLUENCE OF ROLLING DIRECTION AND SHEAR CUTTING ON THE CHARACTERISTICS OF AlMg2 AS FUEL CLADDING MATERIAL FOR RESEARCH REACTOR FUEL. Gradual rolling process of AlMg2 alloy has been carried out to obtain a certain thickness with final mechanical behavior suitable for fuel cladding of research reactor. The post-irradiation testing of fuel element used mini specimens to minimize radioactive waste. The purpose of this study was to study the influence of rolling direction and shear cutting on AlMg2 alloy characteristics. The tests performed include tensile, metalographic and hardness tests after rolling. The results show that AlMg2 alloy rolled by same direction has lower tensile strength and hardness and is more ductile than that rolled by transverse direction. The structure of AlMg2 alloy after hot rolling consists of  (Al) phase with precipitates distributed along the elongated grains in the rolling direction. Shear cutting affects the width and hardness of the deformation area on the cut surface of the specimens. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7926 10.17146/urania.2020.26.3.6090 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 131-142 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7926/6058 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7927 2024-12-23T01:39:26Z urania:ART

MICROSTRUCTURAL EVOLUTION AND COARSENING KINETICS OF B2-(Fe,Ni)Al IN FERITIC STEEL OF Fe-14Ni-9Al-7,5Cr-1Mo AT HIGH TEMPERATURES. Basuki, Eddy Agus Bachri, Syaiful Prajitno, Djoko Hadi Ferritic steels B2-(Fe,Ni)Al precipitate coarsening microstructure evolution MICROSTRUCTURAL EVOLUTION AND COARSENING KINETICS OF B2-(Fe,Ni)Al IN FERRITIC STEEL OF Fe-14Ni-9Al-7,5Cr 1Mo AT HIGH TEMPERATURES. Increasing the efficiency of power plant facilities, such as steam power plant, is demanding, and this drives to the development of materials capable for higher temperature operations. Ferritic steels have been widely used for boiler components due to their relatively economical and higher in thermal conductivity. Nevertheless, the application of ferritic steels for boilers is limited only at 630 ℃ due to their weakness in creep strength and corrosion at higher temperatures. Efforts to increase the performance of ferritic steels at higher temperature applications have been intensively carried out, one of which is by developing B2-(Fe,Ni)Al precipitates in the ferrite matrix. This paper discusses the microstructural evolution in alloy of Fe-14Ni-9Al-7.5Cr-1Mo (in wt.%) strengthened by B2-(Fe,Ni)Al precipitates during heating at 800, 900, and 1000 oC for 6, 20 and 48 hours. Ageing at 800 oC for 6 hours gave rounded cuboidal B2-(Fe,Ni)Al precipitates dispersed homogeneously in the ferrite matrix and changed to rounded shape at longer ageing times. At 1000 oC, however, the precipitates had rounded shape at all times of ageing. Coarsening of precipitates occurred during ageing at 800 and 1000 oC with the rate constants of 5,7 x 104 nm3/h and 2,1 x106 nm3/h respectively, which is considered relatively low. The highest hardness value of 520 HVN was observed for the sample aged at 1000 °C for 6 hours. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7927 10.17146/urania.2020.26.3.6043 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 143-154 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7927/6059 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7937 2024-12-23T01:39:26Z urania:ART

ANALISIS KOMPOSISI UNSUR, DENSITAS, MAKROSTRUKTUR, DAN FASA PADUAN U-6Zr-xNb PASCA UJI KOROSI Masrukan, Masrukan Mustika, Deni Perdana, Deninta Andara Jumaeri, Jumaeri UZrNb alloy elemental composition impurities corrosion density macrostructure phase formation ANALYSIS OF ELEMENTAL COMPOSITION, DENSITY, MACROSTRUCTURE, AND PHASE FORMATION OF U-6Zr-xNb ALLOY POST CORROSION TEST. Research on the elemental composition, density, macrostructure, and phase formation of U-6Zr-xNb alloy have been done after corrosion test. The alloy composition analysis had been done before the corrosion test which included uranium content analysis by potentiometric titration, impurity analysis with Atomic Absorption Spectroscopy (AAS), and U, Zr and Nb content analysis with X-Ray Fluorescence (XRF). The alloy composition analysis was aimed to ensure that U-6Zr-xNb fuel meets the quality requirements of nuclear fuel specifications. The density, macrostructure and phase test were done after the corrosion test. The density test is required as one of the parameters in calculating the corrosion rate, and the microstructure and alloy phase observations are required to determine the damage or corrosion products and the layers/phases formed after the occurance of corrosion. The results of the U-6Zr-xNb alloy composition analysis with XRF and potentiometric titration show that the uranium content is close to the expected level, whereas the impurity test with AAS shows high levels of impurities including Al, Fe and Si. Based on these results, U-6Zr-xNb fuel meets the requirements because the elements of Al, Fe and Si have low neutron absorption cross-section. The sample density test results of U-6Zr, U-6Zr-1Nb, U-6Zr-4Nb, and U-6Zr-7Nb were 16.9798 g/mL, 16.6115 g/mL, 15.594 g/mL, and 15.3564 g/mL respectively. Macrostructure observation of the post-corrosion alloy shows black spots which are the result of oxidation of U (IV) to U (VI). The U-6Zr-xNb alloy experiences the greatest corrosion in mineral-free water media. The damage to the alloy surface decreases as the weight percentage of Nb in the alloy increases. The results of post-corrosion alloy characterization using XRD show that the U-6Zr and U-6Zr-1Nb samples formed an α phase, while the U-6Zr-4Nb and U-6Zr-7Nb samples formed γ phase. The protective Nb2O5 oxide layer formed, however, is minute and unlikely to be detected by XRD. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7937 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 155-166 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7937/6065 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7938 2024-12-23T01:39:26Z urania:ART

PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION Mustika, Deni Torowati, Torowati Dimyati, Arbi Sudirman, Sudirman Fisli, Adel Joni, I Made Langenati, Ratih Indonesian natural graphite purification nuclear fuel matrix acid leaching chemical characterization PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION. Graphite matrix in Pebble Bed Reactor (PBR) – High Temperature Gas Cooled Reactor (HTGR) has an important role as heat transfer medium, neutron moderator and structural material to protect fuel. Thus, graphite matrix must fulfill chemical and physical characteristics for PBR-HTGR fuel. Indonesia has graphite sources in several regions that can potentially be purified. This research aimed to purify Indonesian natural graphite by several variation of acids and to perform chemical characterizations. Natural graphite from flotation process was purified by several variations of acid, i. e., hydrofluoric acid (HF), sulphuric acid + nitric acid (H2SO4 + HNO3) and hydrofluoric acid + hydrochloric acid + sulphuric acid (HF + HCl + H2SO4) and subsequently followed by chemical characterizations such as purity level, ash content, and boron quivalent. The highest purity was obtained in the purification process by HF with carbon content up to 99.52%; this purity level fulfills the specification of nuclear graphite (>99%). Ash content analysis shows a value in compliance with the specification requirement, i.e., < 100 ppm, and boron equivalent value also fulfills the specification value of < 1 ppm. It can be concluded from this study that the graphite purified by acid leaching with HF can be used as fuel matrix candidate but is qualified as low quality. Futher research is required to produce high quality nuclear graphite, particularly research in the minimization of the impurity by evaporation at temperatures over 950 oC to by far lower the ash content. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7938 10.17146/urania.2020.26.3.6059 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 167-176 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7938/6066 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7939 2024-12-23T01:39:26Z urania:ART

PROSES ELEKTRODEPOSISI UNTUK PENGUKURAN ISOTOP 242Pu DENGAN SPEKTROMETER ALPHA Boybul, Boybul Yanlinastuti, Yanlinastuti Nugroho, Arif Kriswarini, Rosika Ginting, Aslina Br. Electrodeposition lutonium isotop current time alpha spectrometry OPTIMIZATION OF ELECTRODEPOSITION PROCESS FOR 242Pu MEASUREMENT BY ALPHA SPECTROMETRY. In this study electrodeposition method was used to design and set up an electrodeposition device for 242Pu source preparation for measurement by alpha spectrometry. The purpose of this study was to determine the optimal electrodeposition process parameters so as to obtain 242Pu alpha sources with good spectrum properties and maximum result from source preparation by electrodeposition process. Standard samples of 242Pu with specific activity were prepared from plutonium standard aqueous solution. Experiments were conducted to determine the main parameters that have significant influence on the deposition efficiency. The electrodeposition process parameters analyzed were the effects of time, electric current and anode-cathode distance using an electrolyte solution of ammonium sulphate at pH 3.5. The results of the optimization of the electrodeposition process were optimal time of 2.5 hours, electric current of 1.4A and cathode- anode distance of 10 mm. These parameters' values are the best conditions for the deposition of 242Pu. Alpha measurement results of the electrodeposited source of 242Pu also show good spectrawith electrodeposition process reproductibility of 95.3%. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7939 10.17146/urania.2020.26.3.6071 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 177-188 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7939/6067 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7940 2024-12-23T01:39:26Z urania:ART

ANALISIS ISOTOP 137Cs DALAM PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI DENGAN METODE PENGENDAPAN CHLOROPLATINATE DAN KOLOM PENUKAR KATION RESIN DOWEX Nugroho, Arif Boybul, Boybul Indaryati, Sutri Haryati, Iis Kriswarini, Rosika Noerpitasari, Erlina U3Si2/Al chloroplatinate Dowex resin 137Cs SEPARATION OF CESIUM IN IRRADIATED U3Si2/Al FUEL PLATE WITH A DENSITY OF 4.8 gU/cm3 USING CHLOROPLATINATE DEPOSITION METHOD AND DOWEX RESIN CATION EXCHANGE COLUMN. Dissolution of the U3Si2/Al plate was carried out in hot cell 109 using 6 M HCl and 6 M HNO3 solutions. The separation of cesium and uranium in uranyl nitrate solution was carried out using chloroplatinate deposition method and a cation exchange column. In the chloroplatinate deposition, the process of removing elements that will interfere with the deposition process was performed first, then the separation of heavy elements such as uranium or plutonium was carried out by depositing the 137Cs in the form of Cs2PtCl6. The isotope content of 137Cs in the Cs2PtCl6 deposition was measured with a γ spectrometer. In the Dowex resin cation exchange column method, the uranyl nitrate solution was pipetted 150 μL to be added to 2 mL of aquadest. To this solution was then added 20 μL of Cs carrier and 1 mL of 12 M HCl. The mixture of the solution was used as feed and put into the anion exchange column containing R-Cl resin. The effluent exiting the anion exchange column was fed into the cation exchange column containing R- NH4+ resin. The 137Cs isotope bound to R NH4+ resin in the column was then eluted using 1 M HCI. The effluent was compressed until a volume of ± 2 mL was obtained. The isotope content of 137Cs in U3Si2 / Al plate obtained as a result of separation using the chloroplatinate method for middle slices coded M has an average of 0.000557 g/g fuel and 0.000652 g/g fuel for the bottom slices coded B, while the Dowex resin cation exchange column method gave an average of 0.000747 g/g fuel for middle slices coded M, and 0.000934 g/g fuel for bottom slices coded B. The recovery of 137Cs separation by the anion exchange column method was greater than that of the Cs2PtCl6 deposition method. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7940 10.17146/urania.2020.26.3.6055 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 189-202 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7940/6068 Copyright (c) 2020 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7941 2024-12-23T01:51:56Z urania:ART

LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR Simbolon, S. Susilowati, S.R ADUN sol-solution ADU-gels gelation spherical pheroid LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR. The aim of this research is to produce thousands of microsphere ADU (Ammonium Diuranate) gels by using external gelation for an intermediate HTGR (High Temperature Gas-cooled Reactor) nuclear fuel in laboratory-scale. Microsphere ADU gels were based on sol-solution which was made from a homogeneous mixture of ADUN (Acid Deficient Uranyl Nitrate) which was containing uranyl ion in high concentration, a water soluble organic compound PVA (Polyvinyl Alcohol) and THFA (Tetrahydrofurfuryl Alcohol). The simple unified home made laboratory experimental machine was developed to replace test tube experiment method which was once used due to a tiny amount of microsphere ADU gels produced. It consists of four main parts: tank filled sol-solution connecting to peristaltic pump and vibrating nozzle, preliminary gelation and gelation column. The machine has successfully converted 150 mL sol-solution into thousands of drops which produced 120 - 130 drops in each minute in steady state in ammonia gas free sector. Preliminary gelation reaction was carried out in ammonia gas sector where drops react with ammonia gas in a bat an eye followed by gelation reaction in column containing ammonia solution 7 M. In ageing process, ADU gels were collected and submerged into a vessel containing ammonia solution which was shaken for 1 hour in a shaker device. Isopropyl alcohol (90%) solution was used to wash ADU gels and a digital camera was used to measured spherical form of ADU gels. Diameters in spherical spheroid form were found between 1.8 mm until 2.2 mm. The spherical purity of ADU gels were 10% - 20% others were oblate, prolate spheroid and microsphere which have hugetiny of dimples on the surface. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7941 10.17146/urania.2015.21.1.2254 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-8 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7941/6069 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7942 2024-12-23T01:47:09Z urania:ART

KARAKTERISASI LAPISAN PENYERAP DAPAT BAKAR PADA PERMUKAAN PELET UO2 + DOPAN TiO2 Sungkono Atmono, Tri Mardji Surface layer Burnable absorber UO2 pellet Microstructures Microhardness Chemical composition Crystal structure Burnable absorber layer on the surface of UO2 + doped TiO2 pellets have successfully created using RF sputtering methods. The objective of this research is to obtain of microstructure characters of UO2 + doped TiO2 pellets, thickness, micro hardness, chemical composition and crystal structure of burnable absorber layer on the surface of UO2 pellets. The methods used are the microstructure and layer thickness using optical microscopy, layer hardness with micro Vickers hardness method, chemical composition by XRF spectrometry, and crystal structure by X-ray diffraction. The results showed that the larger of TiO2 content in the pellets then the greater of the grain size in the microstructure of the pellets and the thicker of the layer formed on the surface of UO2 pellets. The hardness of surface layer of UO2 + doped TiO2 sintered pellets are equal and does not depend on the dopant concentration of TiO2. The surface layer of UO2 + 0.3 % TiO2, UO2 + 0.5 % TiO2 and UO2 + 0.7 % TiO2 sintered pellets are containing zirconium respectively 1.97 mg, 2.47 mg and 4.81 mg. Burnable absorber layer on the surface of UO2 + doped TiO2 sintered pellets have ZrB2 phase with a hexagonal crystal structure. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7942 10.17146/urania.2016.22.3.3182 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 133-142 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7942/6070 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7943 2024-12-23T01:47:09Z urania:ART

PEMISAHAN DAN ANALISIS 137Cs DARI LARUTAN PELAT ELEMEN BAKAR U-7%Mo/Al Anggraini, Dian Noviarty, Noviarty Yanlinastuti, Yanlinastuti Ginting, Aslina Br. Kriswarini, Rosika Nugroho, Arif Boybul, Boybul Cesium separation Precipitation method Cation exchange Zeolit Lampung U-7% Mo/AL fuel plate The aim of this research is to get a valid method of separating cesium from U-7Mo/Al fuel plate solution through determination of parameter of method (accuracy, precision, and recovery). Precipitation method and cation exchange method that are used refer to standard ASTM 690-000 and research result of U3Si2/Al. Parameter method determination has been done by using 150 μL sample (U-7%Mo/Al fuel plate solution, 50 μL of standard solution in 2 mL of HCl 0,1 N. The sample solution was undergone precipitation process by using HClO4 concentrated and 225 mg of CsNO3 as carrier in temperature 0oC for an hour, while exchange cation process was done by using 400 mg of resin zeolite Lampung. The analysis of 137Cs in CsClO4 and 137Cs - zeolite was done by gamma spectrometer. Determination of recovery method was done by comparing count value of 137Cs before and after separation process. Recovery of precipitation method was obtained 95.56 % with accuracy and precision measurement of 0.375 % and 1.875 % respectively, while recovery of cation exchange method obtained 26.73 %. To sum up, the results show that precipitation method better than exchange cation method for separation 137Cs from U-7Mo/Al fuel plate solution. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7943 10.17146/urania.2016.22.3.3183 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 143-154 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7943/6071 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7944 2024-12-23T01:51:56Z urania:ART

INTERAKSI BAHAN BAKAR U3Si2-Al DENGAN KELONGSONG AlMg2 PADA ELEMEN BAKAR SILISIDA TMU 2,96 gU/cm3 PASCA IRADIASI Ginting, Aslina Br. Ajiriyanto, Maman Kartaman Supardjo, Supardjo Interaction U3Si2-Alfuel elements plate AlMg2 cladding pre and post irradiation INTERACTION OF U3Si2-Al FUEL ELEMENT WITH AlMg2 CLADDING ON POST IRRADIATION WITH LOADING OF URANIUM 2.96 gU/cm3. Interaction of U3Si2-Al fuel element with AlMg2 cladding on post irradiation of 2.96 gU/cm3 loading of uranium (TMU) of U3Si2-Al fuel elements plate (PEB) has been analyzed. The purpose of this research is to study the changes of microstructure of nuclear fuel elements during iradiation in reactor core. Understanding on interaction of U3Si2-Al fuel meat with AlMg2 cladding onpre and post irradiation needed to study the influence of radiation on fuel elements plate. PEB U3Si2-Al with 2.96 gU/cm3 by size 10 × 10 mm were heated in DTA (Differential Thermal Analysis) furnace with temperature variation at 450, 550, 650, 900 and 1350oC to perform pre irradiation test.Post irradiation samples were cut by size 2 × 10 mm as many as three samples taken from bottom, middle, and top of PEB in hotcell. The metallography preparation for each pieces of pre and post irradiation samples of U3Si2-Al fuel elements platewith 2.96 gU/cm3 weredone through steps mounting, grinding, polishing, and etching.Scanning Electron Microscope (SEM-EDS) were used to observe the pre irradiation microstructure of fuel elements U3Si2-Al with AlMg2 cladding interaction, while the post irradiationmicrostructure were observed by optical microscope in hot cell. The result show the interaction of U3Si2 with Al matrix or AlMg2 cladding in pre irradiation PEB U3Si2-Aloccurred agglomeration formed new compouds of U(Al,Si)x and UAlx formation. Agglomeration formation on heated pre irradiation samples were bigger while heating temperature increased. The post irradiation sampels shoed the oxide layer were formed outside the AlMg2 cladding and the inner side of caldding that contact to the fuel meat formedlight-grey U(Al,Si)xlayer at 1-3 micron of thickness. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7944 10.17146/urania.2015.21.1.2255 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 9-18 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7944/6072 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7945 2024-12-23T01:47:09Z urania:ART

PEMISAHAN Zr – Hf SECARA SINAMBUNG MENGGUNAKAN MIXER SETTLER Biyantoro, Dwi Isyuniarto Masrukan Separation of Zr Hf Extraction Mixer settlers X-ray fluorescence NAA Separation of Zr - Hf continuously using mixer settler 16 stage has been done. The feed solution is zircon nitrate concentration of Zr = 30786 ppm  and Hf = 499 ppm. As the solvent used extractant 60 % TBP in 40 % kerosene. Nitric acid solution used srubbing 1 M. The feed entered into stage to 5 is contacted with solvents direction on the stage to 16 and the scrubbing solution enter the stage to 1. The purpose of this study is to separate Zr and Hf of the results from the process of zircon sand using solvent TBP using 16 stage mixer settlers. Analysis of the feed and the results of the separation process for zirconium (Zr) using X-ray fluorescence instrument which hafnium (Hf) using Neutron Activation Analysis (AAN). Parameter study done of acidity variation of nitric acid in the feed and time variation in various stirring speed. From the research the separation of Zr-Hf, the optimum conditions in acidity feed 4 N HNO3, equilibrium was received after 3 hours, and stirring speed of 3300 rpm obtained extract of zircon (Zr) concentration = 28577 ppm (efficiency of Zr = 92,76 %) with impurities of hafnium (Hf) = 95 ppm. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7945 10.17146/urania.2016.22.3.3184 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 155-166 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7945/6073 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7946 2024-12-23T01:51:56Z urania:ART

ANALISIS TEKSTUR ZIRCALOY-4 MENGGUNAKAN METODE WILLIAM–IMHOF–MATTHIES-VINEL (WIMV) Priyanto, Tri Hardi Li, Mei Juan Zircaloy-4 WIMV pole figure neutron diffraction TEXTURE ANALYSIS OF ZIRCALOY-4 USING WILLIAM-IMHOF-MATTHIES-VINEL (WIMV) METHOD. Zircaloy-4 as a nuclear fuel cladding has been widely examined macroscopically and microscopically, but research using neutron diffraction technique, especially the characterization of crystallite orientation (texture) in Indonesia is still rare. The purpose of this study is to apply the methods of William-Imhof-Matties-Vinel (WIMV) on the characterization of zircaloy-4 material texture. WIMV is one of data analysis method for textures materials obtained by neutron or x-ray diffractions technique. The analysis is obtained that 002 pole figure has the largest texture index (F2) compared to the others pole figures, that are 100, 101, 102 and 110. Index texture (F2) of pole figure 002 for incomplete pole figures, F2 = 4.69 m.r.d with a reliability factor RP0 = 3.28%, and recalculated pole figures, F2 = 4.42 m.r.d with RP1 = 2.97%. For all pole figures, it is obtained that a reliability factor of the average RP0= 6.60%, RP1 = 5.02%, entropy = - 0.5871, and texture index (F2) = 2.34 m.r.d. Data analysis showed that the WIMV method can be used to determine the direction of the crystallite orientation of the pole figure where the most powerful texture leads to the <001> (c-axis direction) in a hexagonal structure. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7946 10.17146/urania.2015.21.1.2256 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 19-28 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7946/6074 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7947 2024-12-23T01:47:09Z urania:ART

USE OF MEMBRANE EMULSION SPAN 80 AND TOPO IN URANIUM EXTRACTION AND STRIPPING Basuki, Kris Tri Nurimaniwathy Puspita, Dian E.H.B, Bambang Membrane emulsion Extraction Stripping Span 80 Kerosene Uranium Membrane emulsion span 80 and TOPO used in uranium extraction and stripping has been done. The extraction was carried outby emulsion membrane H3PO4 in TOPO-Kerosene. The feed or external aqueous phase was uranium in HNO3. The emulgator span-80 was used to obtain a stable emulsion membrane system. The influence factors were percentage of TOPO-Kerosene, time extraction, molarity of external aqueous phase and molarity of internal aqueous. After the emulsion membrane was formed, the extraction and stripping process was performed. The ratio volume feed: volume membrane phase equal to 1 : 1 and volume of 5 % TOPO-Kerosene: Volume 3 M H3PO4 equal 1 : 1 were used. The relatively good yield were obtained at concentration of TOPO in Kerosene and 3 M H3PO4 was 5 %, molarity of internal aqueous phase equal to 1 M, molarity of external aqueous phase 3 M H3PO4 and time extraction equal to 10 minutes with the speed of emulsification was 8000 rpm. At this condition the extraction efficiency of uranium obtained was 97.8 %, the stripping efficiency 52.56 %, and the total efficiency was 53.80 %. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7947 10.17146/urania.2016.22.3.3185 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 167-178 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7947/6075 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7948 2024-12-23T01:51:56Z urania:ART

KARAKTERISTIK PERMUKAAN SERAT SILIKON KARBIDA HASIL PEMINTALAN LISTRIK DARI POLYCARBOSILANE DALAM N,N-DIMETILFORMAMIDA (DMF)/ TOLUENA Mustika, Deni Sihombing, Riwandi Pribadi, Slamet Langenati, Ratih Sujatno, Agus Dimyati, Arbi Setiawan, Jan Indarto, Eddy SiC fibers polycarbosilane toluene DMF morphology SEM surface area pore size nnuclear material composit SiC SURFACE CHARACTERISTIC RESULT OF ELECTRIC SPINNING FROM POLICARBOSILANE IN THE DIMETILFORMAMIDA (DMF)/ TOLUENA. Non oxide ceramic silicon carbide (SiC) has unique properties such as mechanical strength, chemical and thermal stability that is used in many applications. Results of modelling studies on SiC show good stability on neutron irradiation and had low permeability againts fission product increased the interest in the use of SiC in nuclear applications. Composites form was used to increased mechanical properties. Composites that used fiber as filler show better mechanical properties than filler in form particle and whisker. In SiC composites, mechanical properties influenced by interphase properties and/or SiC surface characteristic. Electrospinning was a promising method to form fibers. This research aims to study the characteristics of silicon carbide fiber surface produced from polycarbosilane that solute by toluene and N,N-dimethylformamide (DMF). The addition of DMF and polycarbosilane in toluen affect the electrospinnability and characteristic the surface of fibers. The SiC fibers produced from polycarbosilane that solute by toluen and N,N-dimethylformamide (DMF) form continue fibers with ribbon liked. Black spot on the surface of pyrolysis fibers possible from free carbon or graphite contamination from furnace material. Value of surface area of pyrolysis fibers at 3.321 – 46.14 m2/g and porous in micro size, with distribution its radius in range at 1-3 nm, highgest distribution in size less than 2 nm. Pyrolysis and sintering in higher temperature expected to form fiber with less porous and its density able to approach the theoretical density. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7948 10.17146/urania.2015.21.1.2257 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 29-38 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7948/6077 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7949 2024-12-23T01:47:09Z urania:ART

PENGARUH REGENERASI KOLOM ALUMINA ASAM TERHADAP RECOVERY DAN KUALITAS 99mTc HASIL EKSTRAKSI PELARUT MEK DARI 99Mo HASIL AKTIVASI NEUTRON Gunawan, Adang Hardi Sugiharto, Yono Widyastuti 99mTc 99Mo MEK Acidic alumina column Radiochemical purity Purification of 99mTc from 99Mo activation using acidic alumina column chromatography system from MEK (Methyl Ethyl Keton) extraction has been carried out through cooperation between PTRR - BATAN, Chiyoda and JAEA Japan. This method has a limitation that acidic alumina column for purification of 99mTc can be used only once, for the next purification acidic alumina column should be replaced with new column, so it is less practical and also requires high cost. This study aims to obtain a 99mTc solution can be used for labelling of a radiopharmaceutical kit. In this study, the used of acidic alumina column for 99mTc purification was tried more than once by regeneration using 0.1N HNO3 solution after purification process is completed. Parameters observed in this study are the percent recovery, elution profile, pH, radiochemical purity and radionuclide purity. The results of observational studies conducted over 5 days has been obtained pH ~ 5, % recovery > 60%, radiochemical purity of > 95% and 99Mo leakage not detected. The treatment of acidic alumina column with 0,1 N HNO3 solution concluded that acidic alumina column does not need to be replaced every day. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7949 10.17146/urania.2016.22.3.3186 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 179-188 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7949/6076 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7950 2024-12-23T01:47:09Z urania:ART

REDUKSI AKTIVITAS URANIUM DALAM LIMBAH RADIOAKTIF CAIR MENGGUNAKAN PROSES ELEKTROKOAGULASI Prayitno Ridantami, Vemi Prayogo, Imam Electrocoagulation Reduction of uranium waste Voltage Aluminium Waste generated from the process of the industrial material development one of which waste containing uranium radioactive, can have negative impact on humans and the environment. In the present time, chemicals are still mostly use in radioactive waste treatment. To reduce pollutants with the use of chemicals is less efficient, because less environmentally friendly, take long time and costly. Therefore, a system of electrocoagulation process will be applied to decrease the activity of uranium from waste solution. The purpose of this study is to determine the efficiency of uranium activity decrease in waste water which is produced in the electrolysis process is conducted with voltage variations, the dwelling time, electrode spacing, and the waste inlet pH. The waste that will be treated has uranium contaminant levels of 500 mg/L. The experiment was conducted by a batch system with aluminum electrodes. Parameters affecting electrocoagulation process, such as voltage, time, distance electrode, and the pH have been studied and the best voltage optimization condition has been obtained of 12.50 V, a distance of 1 cm, pH 7, and in the processing time of 60 minutes efficiency of 97.20 % was obtained. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7950 10.17146/urania.2016.22.3.3187 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 189-202 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7950/6079 Copyright (c) 2017 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7951 2024-12-23T01:51:56Z urania:ART

MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR Suwardi, Suwardi Dewayatna, Winter Sungkono, Sungkono Ridwan, Ridwan Rifai, M irradiation fuel pin natural UO2 geometry stress-strain MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*). The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology) then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility) of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7951 10.17146/urania.2015.21.1.2259 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 39-46 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7951/6078 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7952 2024-12-23T01:50:28Z urania:ART

PENGARUH SERBUK U-Mo HASIL PROSES MEKANIK DAN HYDRIDE – DEHYDRIDE – GRINDING MILL TERHADAP KUALITAS PELAT ELEMEN BAKAR U-Mo/Al Supardjo, Supardjo Kadarjono, Agoeng Boybul, Boybul U-7Mo Alloy U-7MO Powder U-7Mo/Al fuel core U-7Mo/Al fuel plates dispersion fuel INFLUENCE OF U-Mo POWDER BY MECHANICAL AND HYDRIDE - DEHYDRIDE - GRINDING MILL PROCESS RESULT OF U-Mo / Al FUEL PLATE QUALITY. Research of U-7Mo/Al fuel type plate is done in order to develop U3Si2/Al fuel to get a new fuel that has a higher uranium density, stable for use as fuel in the reactor and is easily done if the reprocessed. The scope of the research includes manufacture: U-7Mo alloy with smelting techniques, pulverizing U-7Mo to be filed and hydride–dehydride–grinding mill, U-7Mo/Al fuel core with the technique of compacting at a pressure of 20 bar, and U-7Mo/Al fuel plate with technique of hot rolling at a temperature of 425oC. The U-7Mo alloy results smelting process quite homogeneous, the density of 16.34 g/cm3 and is tenacious, then made powder by means of filed and hydride–dehydride–grinding mill. The U-7Mo powder shaped flat results miserly process, contaminants Fe is high enough, whereas powder process results hydride- dehydride-grinding mill, tend equiaxial with low contaminants. The second type of U-7Mo powder is used as a raw material for making U-7Mo/Al fuel core and U-7Mo/Al fuel plate with 7 gU/cm3 uranium density and obtained product with almost the same quality. The U-7Mo/Al fuel core test results measuring 25 x 15 x 3.15 ± 0.05 mm, there is no defect/crack, U-7Mo distribution in the matrix is quite homogeneous and there is no grouping/agglomeration U-7Mo dimension >1 mm. The U-7Mo/Al fuel plate outcome rolling with a final thickness of 1.45 mm, has a thickness of 0.60 mm and a mean meat cladding thickness of 0.4 mm, and there is one point of measurement of cladding with a thickness of 0.15 mm. By comparing the use of both types of U-7Mo powders the U-7Mo/Al fuel core and U-7Mo/Al fuel plate produced has almost the same quality. However, the use of U-7Mo powder results hydride– dehydride–grinding mill process is better because the workmanship is faster and impurities in the powders can be minimized. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7952 10.17146/urania.2015.21.2.2260 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 47-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7952/6084 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7958 2024-12-23T01:50:28Z urania:ART

EFFECT OF Nb ELEMENT CONTENT IN U-Zr ALLOY ON HARDNESS, MICROSTRUCTURE AND PHASE FORMATION Masrukan, Masrukan Al Hasa, M. Husna Setiawan, Jan Pribadi, Slamet Nb element U-Zr alloy hardness microstructure phase analysis EFFECT OF Nb ELEMENT CONTENT IN U-Zr-Nb ALLOY ON HARDNESS, MICROSTRUCTURE AND PHASE FORMATION. Experiments to determine the effect of Nb element in the U-Zr alloys on hardness, microstructure and phase formation has been done. The addition of Nb element would effect the hardness, microstructure and phase which formed. The U-Zr-Nb alloy was made with the variation of Nb 2%, 5% and 8% by melting in an electric arc melting furnace that equipped with water cooling and the argon atmosphere. The U-Zr-Nb alloy to be cut divided to some testing, such as hardness test, microstructure, and phase analysis. Hardness testing was done by Vickers hardness testing equipment, microstructure by an optical microscope, and diffraction pattern by XRD and phase analysis was done by GSAS. Hardness testing results showed that the addition of 2% to 5% Nb element in U-Zr alloys will increased in hardness, but the addition of Nb element over 5% the hardness was decreased. Observations the microstructure showed that the addition of 2% to 5%Nb element, grains were formed from fine into coarse. Phase analysis for diffraction pattern showed that the phase changed from αU and γU (Zr,Nb)at 2% Nb to be αU, γU (Zr,Nb) and δ1 (UZr2) phase at 5% and 8% Nb. Phase changes was followed by changes in its compositions. The composition of αU at 2% Nb was 40% increased to 81% at 5% Nb and decreased to 3.9% at 8% Nb. The composition of γU decreased from 59,86% to 14,91% with increased Nb from 2% to 5% and further increased to 52,74% at 8% Nb. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7958 10.17146/urania.2015.21.2.2261 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-64 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7958/6085 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7959 2024-12-23T01:50:28Z urania:ART

KARAKTERISTIK MIKROSTRUKTUR DAN FASA PADUAN Zr- 0,3%Nb-0,5%Fe-0,5%Cr PASCA PERLAKUAN PANAS DAN PENGEROLAN DINGIN Sungkono, Sungkono Zr-0.3%Nb-0.5%Fe-0.5%Cr alloy heat treatment cold rolling microstructure phase MICROSTRUCTURE AND PHASE CHARACTERISTICSOF Zr-0.3%Nb-0.5%Fe-0.5%Cr ALLOY POST HEAT TREATMENT AND COLD ROLLING. Zr-Nb-Fe-Cr alloys was developed as fuel elements cladding with high burn up for advanced power reactors. In this research has been made of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy were heat treated with varying temperatures at650 and 750°C for 1 until 2 hours. The objectives of this research was to obtain the character of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy post heat treatment and cold rolling, microstructure nomenclature, crystal structure and phases that presents in the alloy. The results of this experiment showed that the microstructures of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy post heat treatment (650ºC, 1-2 hours) had equiaxial grain structure with an enlarged size with increasing of the retention time. Meanwhile, post heat treatment (750°C, 1-2 hours) occurred the microstructures evolution of alloy from equiaxial and columnar became equiaxial and columnar relatively large, and subsequently became the larger equiaxial grains. Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy post heat treatment (650°C, 2 h) and (750°C, 1.5-2 hours) can undergo cold deformation without cracking when thickness reduction between 5 to 10%.The phases formation of Zr2Fe, ZrCr2 and FeCr compounds can improve the mechanical strength and the corrosion resistance of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7959 10.17146/urania.2015.21.2.2262 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 65-76 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7959/6086 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7960 2024-12-23T01:50:28Z urania:ART

KETAHANAN KOROSI PADUAN Al-Mg 5052 DI DALAM AIR PENDINGIN NETRAL MENGANDUNG KLORIDA Jatmiko, Dicky Tri Nurdin, Isdiriayani Devianto, Hary Al-Mg Alloy 5052 chloride corrosion pH temperature water CORROSION RESISTANCE OF Al-Mg ALLOY 5052 IN CHLORIDE CONTAINING NEUTRAL COOLING WATER. Al-Mg alloy 5052 is a material used as nuclear fuel element cladding due to its low neutron flux absorption and high corrosion resistance in demineralized water. This research is focused to know of the corrosion resistance of Al-Mg alloy 5052 in chloride containing neutral water used as demineralized primary cooling water substitute in GA Siwabessy Multi Purpose Reactor (RSG-GAS). This research covers the corrosion rate measurement using the Tafel method, corrosion process prediction using cyclic voltammetry method and corrosion product analysis using X-Ray Diffraction method. The experiments are carried out at temperature variation of 30°C, 35°C, 40°C and 45°C, as well as sodium chloride concentration of 0.05 M, 0.25 M and 0.5 M. The research results show that Al-Mg alloy 5052 is insignificantly to moderate corroded in chloride containing water by single step irreversible oxidation reaction. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7960 10.17146/urania.2015.21.2.2263 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 77-86 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7960/6087 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7961 2024-12-23T01:50:28Z urania:ART

OPTIMASI PENGGUNAAN HCl SEBAGAI LARUTAN PENGELUSI ITRIUM-90 DALAM DOWEX® 50WX8-200 Sulaiman, Sulaiman Aguswarini, Sri Karyad, Karyad Chairuman, Chairuman S, Gatot Subur, M H G, Adang yttrium 90 Y-citrat 90 YCl3 Dowex® 50WX8-200 generator 90 Sr/ 90 Y HCl OPTIMIZATION OF USE HCl AS SOLUTION FOR ELUTION ITRIUM-90 IN DOWEX® 50WX8- 200. Yttrium-90 is a β emitter radionuclide with a half life of 64.1 hours and emits the maximum β energy at 2280 keV. Yttrium-90 radionuclide is widely used in nuclear medicine for therapeutic purposes. 90Sr/90Y generator system has been developed to obtain carrier-free 90Y. Yttrium-90 were obtained from the 90Sr/90Y generator in the form of 90Y-citrate radiochemical. The 90Y ideal for tagging ligands is in the form of 90YCl3. To obtain the desired radiochemical a series of processes with ion exchange resin method is performed using Dowex® 50WX8-200. In this study optimization of the use of HCl for has been carried out 90Y elution from the resin with varied HCl concentration and contact time when 90Y is eluted from the resin. The release of 90Y from resin Dowex® 50WX8-200 depending on the concentration of HCl is used. The greater the concentration of HCl release rate of 90Y from 90Y-Dowex greater. The optimum contact time with 5 mL of 12 N HCl is 60 minutes with a result of 9.4%. The 90Y test results after eluted from Dowex® 50WX8-200 indicated that the 90Y was in the form of 90YCl3. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7961 10.17146/urania.2015.21.2.2264 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 87-94 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7961/6088 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/7962 2024-12-23T01:49:44Z urania:ART

PEMBUATAN SERBUK U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET Masrukan, Masrukan Sungkono, Sungkono Yanlinastuti, Yanlinastuti Yulianto, Tri Ridwan, Ridwan U-6Zr powder U enrichment of 19.75 % fuel research reactor Making U-6Zr alloy powder with enrichment of 19.75 % for a research reactor fuel has been done. Making of U-6Zr fuel in order to find new fuels that have a high density to replace the existing fuel U3Si2-Al. The purpose of this experiment was to determine the properties of U-6Zr alloy powder obtained from the hydriding-dehydriding process as a candidate research reactor fuel. Initially was made U-6Zr ingot by melting U and Zr metals using electric arc melting furnace. U-6Zr ingots were found then cut into pieces and put into hydriding-dehydriding equipment that operates at pressures 14,46054 psi and hydriding temperatures of 350 oC. Ingots that out of the hydriding-dehydriding equipment becomes brittle then was crushed so becomes powder. Powder that obtained subjected to the test, including chemical composition testing, density, hydrogen content, phase and thermal properties. The results testing show that chemical composition of some elements such as Al, Ca, Cu, and Ni exceeded the permitted limit, each of which contained elements of 202.21 ppm, 214.05 ppm, 61.25 ppm, and 134.53 ppm. On testing the density obtained that U-6Zr powder density of 13.57 g/cc and the testing of hydrogen residual content obtained hydrogen residual content of 0.16 %. For the testing phase were obtained the αU and δU phases, while in testing of the transformation temperature, there are two peaks, the first peak occurs at a temperature of 274 to 311 °C and a second peak occurs at a temperature of 493 to 527 oC. The first peak occurs endothermic reaction by absorbing heat of 6.23 cal/g but not formed a new phase, while the second peak occurs exothermic reaction with brought out heat of -9.34 cal/g and formed αZr phase. Meanwhile, the heat capacity of the test at temperature of 34 to 75 °C, the decrease in the value of the heat capacity accompanied heat absorption. At higher temperatures up to 437 °C, heat capacity value becomes negative and accompanied by heat expenditure. It can be concluded that the alloy the U-6Zr when viewed from the chemical composition can still be used for fuel research reactor. Residual hydrogen still found in small quantities in the U-6Zr powder that can be eliminated by heating for longer. Reaction between Zr with residual hydrogen to form the αZ phase is accompanied exothermic reaction with brought out heat of 93 449 cal/g. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/7962 10.17146/urania.2015.21.3.2460 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 95-104 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/7962/6128 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8007 2024-12-23T01:49:44Z urania:ART

PENGARUH PENGEROLAN PANAS TERHADAP KARAKTER PADUAN Zr-0,6Nb-0,5Fe-0,5Cr Sungkono, Sungkono Aidah, Siti Alloy hot rolling thickness reduction microstructure hardness phase Characters of Zr-0.6Nb-0.5Fe-0.5Cr alloy after hot rolling have been studied. The objective of this research was to obtain of hot rolling effect on the characteristics of microstructures, hardness and phases formed in Zr-0.6Nb-0.5Fe-0.5Cr  alloy. The hot rolling process of alloy carried out at temperature of  800 oC with retention time of 1.5 and 2 hours and a thickness reduction between 5 to 25 %. The results of this experiment showed that the Zr-0.6Nb-0.5Fe-0.5Cr alloy has Widmanstatten structure with microstructure evolving into deformed columnar grains and deformed elongated grains with increasing thickness reduction. Besides, the longer the retention time at temperature of 800 oC is the larger are the grain structures and formation of a-Zr and Zr3Fe phase. The hardness of  Zr-0.6Nb-0.5Fe-0.5Cr alloy has same trends i.e the larger thickness reduction gives higher hardness. The Zr-0.6Nb-0.5Fe-0.5Cr alloy can undergo hot rolling deformation at a thickness reduction of  25 %  and the formation of a-Zr and Zr3Fe can increased of hardness and strength of Zr-0.6 Nb-0.5 Fe-0.5 Cr alloy. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8007 10.17146/urania.2015.21.3.2461 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 105-116 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8007/6129 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8008 2024-12-23T01:49:44Z urania:ART

PENGARUH PEMANASAN KERNEL UO2 DALAM MEDIUM GAS ARGON TERHADAP SIFAT FISIS KERNEL UO2 SINTER Damunir, Damunir Susilowati, Sri Rinanti Dewi, Ariyani Kusuma UO2 kernel heating physical properties density porosity The heating of UO2 kernels in argon gas medium on the physical properties of sinter UO2 kernels was conducted. The heated of the UO2 kernels was conducted in a sinter reactor of a bed type. The sample used was the UO2 kernels resulted from the reduction results at 800 oC temperature for 3 hours that had the density of 8.13 g/cm3; porosity of 0.26; O/U ratio of 2.05; diameter of 1146 µm and sphericity of 1.05. The sample was put into a sinter reactor, then it was vacuumed by flowing the argon gas at 180 mmHg pressure to drain the air from the reactor. After that, the cooling water and argon gas were continuously flowed with the pressure of 5 mPa with 1.5 liter/minutes velocity. The reactor temperature was increased and variated at 1200-1500 oC temperature and for 1-4 hours. The sinters UO2 kernels resulted from the study were analyzed in term of their physical properties including the density, porosity, diameter, sphericity, and specific surface area. The density was analyzed using pycnometer with CCl4 solution. The porosity was determined using Haynes equation. The diameters and sphericity were showed using the Dino-lite microscope. The specific surface area was determined using surface area meter Nova-1000. The obtained products showed tha the heating of UO2 kernel in argon gas medium were influenced on the physical properties of sinters UO2 kernel. The condition of best relatively at 1400 oC temperature and 2 hours time. The product resulted from the study was relatively at its best when heating was conducted at 1400 oC temperature and 2 hours time, produced sinters UO2 kernel with density of 10.14 gr/ml; porosity of 7 %; diameters of 893 µm; sphericity of 1.07 and specific surface area of 4.68 m2/g with solidify shrinkage of 22 %. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8008 10.17146/urania.2015.21.3.2462 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 117-126 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8008/6130 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8009 2024-12-23T01:49:44Z urania:ART

PEMBUATAN SUMBER RADIASI GAMMA 137Cs DENGAN AKTIVITAS 20 mCi DARI PEB U3Si2-Al PASCA IRADIASI DALAM CONTAINER STAINLESS STEEL Ginting, Aslina Br Yanlinastuti, Yanlinastuti Noviarty, Noviarty Boybul, Boybul Nugroho, Arif Anggraini, Dian Kriswarini, Rosika Sriyono, Sriyono Subechi, Moch Wurdiyanto, Gatot Hermawan, Hermawan Process wastle of PEB U3Si2-Al gamma radioactive isotope 137Cs cation exchange zeolite Lampung and container In the post-irradiation examination of fuel element plate (PEB) U3Si2-Al), a solution of high activity as a result of testing nuclear fuel stored in hot cell with enough volume. The solution cannot be discarded as waste because it still contains fission isotope such as137Cs, uranium and trans uranium, which has a long half-life and dangerous for the environment. This can help the industry in order to fulfill the needs of a radioactive source in Indonesia, because until now 137Cs isotope is derived from foreign industries. In addition, the procurement and transportation of isotopes 137Cs require stringent requirements, because they have toget permission from the National Nuclear Energy Agency (BAPETEN), thus causing the price of high activity 137Cs isotopes becomes expensive to Indonesia. For these reasons, BATAN as nuclear R&D institutions in Indonesia need to study make isotopes 137Cs gamma radiation source, which is contained in the waste from spent fuel test results U3Si2-Al. Isotope 137Cs can be used very widely, such as in the analysis of environmental samples, the oil and gas industry, construction, radiography, fisheries, hospitals, and mining. Making isotope 137Cs gamma radiation source starting from the collection of waste from the test results PEB U3Si2- Al. Waste solution was collected in a bottle with volume 65 mL. Collection of 137Cs isotopes of other fission carried out using the method of cation exchange with weight 45 gr of zeolite Lampung. The results of separation are 137Cs-zeolite in the solid phase and the other isotopes are in the liquid phase. 137Cs-zeolite solid is then dried and then weighed and measured its activity using a spectrometer-g. Result of analysis by spectrometer-g was obtained activity of 137Cs-zeolite solids was 20 mCi.137Cs-zeolite solids then packed in sealed containers (shield source) capsule-shaped stainless steel and then certificate by PTKMR-BATAN. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8009 10.17146/urania.2015.21.3.2463 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 127-140 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8009/6131 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8010 2024-12-23T01:49:44Z urania:ART

PENENTUAN KANDUNGAN ISOTOP 235U DALAM PEB U3Si2-Al TMU 2,96 gU/cm3 UNTUK PERHITUNGAN BURN-UP Boybul, Boybul Yanlinastuti, Yanlinastuti Indaryati, Sutri Haryati, Iis Nugroho, Arif PEB U3Si2-Al post-irradiation top 235U anion exchanger and burn-up 235U separation has been carried out in the solution of PEB U3Si2 - Al with loading of uranium (TMU) 2.96 gU/cm3 at the Top. Top of PEB U3Si2-Al cut into three sections (triplo) with code T1, T2 and T3. Weight of each PEB code T1 = 0.095 g, T2 = 0.086 g and T3 = 0.087 g and dissolved using HCl and HNO3 in hotcell. The purpose of this study was to determine the content of the isotope 235U in the solution PEB U3Si2-Al post-irradiation and subsequently used for the calculation of burn-up. 235U isotope separation in the solution PEB U3Si2 - Al was conducted using an anion exchange column method using Dowex1x8 resin. Pipette solution of 100 mL, and then put into a column containing Dowex resin with a weight of 1.2 g. U effluent results in the column was eluted using 0.1 M HCl, then dried and conducted electro-deposition process and then analyzed using a spectrometer-α. The analysis showed that the content of the isotope 235U obtained at T1 = 0.03665 g/g PEB, T2 = 0.003468 g/g PEB and T3 = 0.03208 g/g PEB with separation recovery of 63.71 %. The content of isotope 235U obtained is used for the calculation of burn-up. burn-up calculation results of PEB U3Si2-Al of Top section (T1, T2 and T3) were obtained respectively by 43.31 %, 49.48 % and 45.41 % or burn-up an average of 45.75 %. This data is smaller than a data burn-up of PEB U3Si2-Al of middle section of 50.69 % conducted by previous researchers. However, this data can not be used as an input to the reactor, due to should be equipped with a data burn-up of PEB U3Si2-Al of bottom section. Therefore, in the nextstudies will be conducted calculation of burn-up PEB U3Si2-Al of bottom section. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8010 10.17146/urania.2015.21.3.2464 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 141-150 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8010/6132 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8011 2024-12-23T01:49:44Z urania:ART

FENOMENA KOROSI ZIRKALOY-2 DAN ZIRKALOY-4 DALAM MEDIA LARUTAN NaCl SECARA ELEKTROKIMIA Anggraini, Dian Ajiriyanto, Maman Kartaman Kriswarini, Rosika Yanlinastuti, Yanlinastuti Corrosion zirkaloy-2 and zirkaloy-4 NaCl solvent medium electrochemical polarization technique Zircaloy-2 and zircaloy-4 are used as cladding material in Light Water Reactor, (LWR), Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR). During its use in the spent fuel pool may interact between the cladding and seawater in case of abnormal condition. Chloride ion which contained in seawater has the potential for being corrosion in cladding material. The aim of this work was knowing zircaloy-2 and zircaloy-4 corrosion phenomena in NaCl medium by corrosion parameter determination with polarization resistance technique, tafel, and potentiodynamic. Samples are zircaloy-2 and zircaloy-4 pieces material with 1x1 cm dimension. The samples are mounting and soldered with copper wire, then the sample surface are polished by 1200 grade grinding paper. Corrosion test was done with corrosion cell that was completed by electrode standard (saturated calomel), electrode buffer (graphite) and work electrode (sample). Solvent medium was NaCl solution with concentration 3.5 %; 0.35 %; and 0.175 %. The result showed that ion Clˉ contained in NaCl which were concentration area from 0.175 % to 3.5 % did not affect corrosion mechanism. The corrosion type in potential range ± 250 mV to E­­corr was uniform corrosion in ZrO­2 oxide form. Zirkaloy-2 and zirkaloy-4 corrosion rate in 3.5 % NaCl concentration were obtained 6.39×10-3 and 7.40×10-3 mpy respectively. Corrosion phenomenon which was observed by potentiodynamic technique (± 1000 mV to Ecorr) showed that zirkaloy-2 and zirkaloy-4 underwent passivation and pitting corrosion. Pitting potential zirkaloy-2 and zirkaloy-4 were obtained -452.8 mV and -182.8 mV severally. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8011 10.17146/urania.2015.21.3.2465 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 151-159 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8011/6133 Copyright (c) 2015 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8012 2024-12-23T01:47:58Z urania:ART

FABRIKASI MIKROSFIR UO2 MENGGUNAKAN TEKNIK AERASI Rachmawati, Meniek Mutiara, Etty Yulianto, Tri UO2 Fuel Sol-gel Microsphere Aeration It has been developed a fabrication process of low density UO2 microspheres for direct feed in pelletization process of PHWR advanced reactor fuel. The fabrication has been implemented by a sol-gel method using aeration technique on the sol / broth with external methods and three variations of gelation. In the aeration technique, broth is directly prepared in gelation mode without one night settling time. The aeration technique is the opposite of the deaeration technique, which is used in fabricating microspheres HTGR fuel. The broth which has been prepared with a mole ratio of NO3 / U between 1.5 to 1.7 with pH solution of 1.6 and viscosity between 630-660 cP directly to be gelated in three ways of gelation. The process of both gelation method 1 and method 2 is implemented by passing through the broth to a dispersion nozzle with 1 mm of diameter which is vibrated by an electromagnetic vibrator at 150 Hz with medium for free fall droplet differently before entering into a solution of NH4OH, while the gelation method 3 implemented manually. The wet UO2 microspheres derived from the three ways of gelation above are treated by heating process at the same way that is dried at a temperature of 85 ºC and 220 ºC each respectively for 1 hour, followed by a calcination process of microspheres UO2 for 1 hour at a temperature of 500 ºC in gaseous medium of O2 and a reduction process at a temperature of 600 ºC in gaseous mixture of N2 and H2 medium for 1 hour. The UO2 microspheres gelation of method 3 are chosen to be sorted and characterized. The characterization results provide the characteristics data of UO2 microspheres in the form of microspheres diameter of 900 μm, tap density of 1.90 g/cm3 and specific surface area of 6 m2/g. The results of the analysis and characterization are then compared to other data of other research so that it can be concluded that the use of such an aeration technique on the broth can produce low density UO2 microspheres that qualify as direct feed for fuel pelletization process of the advanced PHWR and advanced PWR. The fabrication using the aeration technique with gelation method 2 has the greatest opportunity to produce UO2 microspheres with a high production rate and certain characteristics if it is implemented by resetting the rate of dispersion and the free fall droplet duration. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8012 10.17146/urania.2016.22.2.3089 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 65-74 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8012/6135 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8013 2024-12-23T01:47:58Z urania:ART

KORELASI KOMPOSISI UNSUR TERHADAP SIFAT TERMAL SERBUK BAHAN BAKAR U-ZrHX Masrukan Alhasa, M. Husna Yanlinastuti, Yanlinastuti Composition Thermal properties Fuels U-ZrHx Analysis has been conducted to determine the composition correlation on the thermal properties of the powder fuel U-ZrHx. U-ZrHx powder made from the process hydriding U-Zr ingot, where the ingot is the result of U-Zr and Zr U metal melting. In this experiment made three variations of powders, namely U-35ZrHx, U-45ZrHx, and U-55ZrHx. Need for determination of the thermal properties of Zr was to determine the effect of the nature of the transformation of Zr levels of heat from the fuel. At first, U and Zr metal is melted in electric arc furnaces to produce ingot U-Zr. U-Zr ingot then made powder with hydridring-milling techniques to produce U-Zr powder. U-Zr powder composition analyzed using techniques spectroscopy atomic absorption (AAS) and UV-Vis spectroscopy. The results of composition analysis showed that the analysis to determine the content of U and Zr nearly all the test samples analyzed have quite a big difference between the content of U and Zr as determined by the results of the analysis of U and Zr exception analysis result in powder U-45Zr which differ only 0.609%. From the analysis of impurities obtained that nearly all the impurities that exist still meet the requirements for fuel unless the elements Fe, where elements of the existing Fe amounted to 382.912 g/g while the requirement of £ 250 mg /µg. Testing conducted heat capacity in the temperature range 35 ° C to 437 ° C showed that the capacity were greatest powder 35ZrHx U-with a value heat capacity of 0.13 J / g.oC. Meanwhile, test results obtained transition phase change that the U-45ZrHx undergo two stages of reaction with phase change. It can be concluded when seen from the content of U and Zr cannot be used for fuel, while the analysis of the content of impurities found that all the elements that are still eligible for the fuel unless the element Fe. For the analysis of the thermal properties are the heat capacity of the heat capacity of the highest values obtained in powder U-35ZrHx, while the transition from the testing phase changes shows that the U-45ZrHx undergo two stages of reaction with phase change. There is an effect of the composition on the thermal properties, where the higher the content of Zr, the value of uranium zirconium hydride heat capacity is lower. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8013 10.17146/urania.2016.22.2.3090 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 75-86 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8013/6137 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8016 2024-12-23T01:47:58Z urania:ART

PHASE CHANGES ON 4H AND 6H SIC AT HIGH TEMPERATURE OXIDATION Setiawan, Jan Suryaman, Ganisa Kurniati Masrukan Silicon carbide 4H-SiC 6H-SiC Oxidation High temperature The oxidation on two silicon carbide contains 6H phase and contains 6H and 4H phases has been done.  Silicon carbide is ceramic non-oxide with excellent properties that potentially used in industry.  Silicon carbide is used in nuclear industry as structure material that developed as light water reactor (LWR) fuel cladding and as a coating layer in the high temperature gas-cooled reactor (HTGR) fuel.  In this study silicon carbide oxidation simulation take place in case the accident in primary cooling pipe is ruptured.  Sample silicon carbide made of powder that pressed into pellet with diameter 12.7 mm and thickness 1.0 mm, then oxidized at temperature 1000 oC, 1200 oC dan 1400 oC for 1 hour.  The samples were weighted before and after oxidized.  X-ray diffraction con-ducted to the samples using Panalytical Empyrean diffractometer with Cu as X-ray source.  Diffraction pattern analysis has been done using General Structure Analysis System (GSAS) software. This software was resulting the lattice parameter changes and content of SiC phases.  The result showed all of the oxidation samples undergoes weight gain.  The 6S samples showed the highest weight change at oxidation temperature 1200 oC, for the 46S samples showed increasing tendency with the oxidation temperature.  X-ray diffraction pattern analysis showed the 6S samples contain dominant phase 6H-SiC that matched to ICSD 98-001-5325 card.  Diffraction pattern on 6S showed lattice parameter, composition and crystallite size changes.  Lattice parameters changes had smaller tendency from the model and before oxidation.  However, the lowest silicon carbide composition or the highest converted into other phases up to 66.85 %, occurred at oxidation temperature 1200 oC.  The 46S samples contains two polytypes silicon car-bide.  The 6H-SiC phases matched by ICSD 98-016-4972 card and 4H-SiC phase matched by ICSD 98-016-4971 card.  Diffraction pattern on 46S also showed lattice parameter, composition and crystallite size changes.  The lattice parameter changes not significant.  For 6S and 46S samples at 1400 oC, the 6H-SiC phase changes into other phases more than 50 % from its original weight percentage. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8016 10.17146/urania.2016.22.2.3091 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 87-98 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8016/6139 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8017 2024-12-23T01:47:58Z urania:ART

METODE PENGENDAPAN DAN PENUKAR KATION UNTUK PEMISAHAN CESIUM DALAM BAHAN BAKAR U3Si2-Al Ginting, Aslina Br. Anggraini, Dian 137Cs isotope Zeaolite CsNO3 powder Cation exchange and precipitation method 137Cs isotope was one of the fission products which could be used as a burn up monitor to study the performance of the fuel during irradiation in a reactor. For analyzed 137Cs isotope in post-irradiation U3Si2-Al fuel element plate (PEB), it was needed a valid method to obtain accurate data. Several methods could be used for 137Cs separation from U3Si2-Al PEB solution were precipitation method in CsClO4 salt form according to ASTM E 320-79 and cation exchange method using Lampung zeolite. The precipitation process was done using CsNO3 powder as a carrier substance and HClO4 reagent, while the cation exchange process was done by the addition of Lampung zeolite. The objective research was getting a valid method for 137Cs separation in post-irradiation U3Si2-Al PEB, especially for the aspects of CsNO3 powder weight influence and weight of zeolite Lampung was added. 137Cs isotope precipitation process was done by pipetting 150 mL U3Si2-Al PEB solution which was added to the 500; 625; 700; and 1000 mg weight variation CsNO3 powder and 4 mL HClO4 in ice media for 1 hour. The precipitation process results were obtained CsClO4 precipitate separated from the supernatant as a liquid phase. Meanwhile, the cation exchange process was done by adding 700; 900; 1000 and 1200 mg weight variation Lampung zeolite by shaking for 1 hour. The cation exchange process results was obtained 137Cs-zeolite in the solid phase and the other isotopes in the liquid phase. The137Cs content in 137CsClO4 precipitate,137Cs-zeolite solid phase and supernatant was measured using spectrometer-g. The analysis showed that the optimum weight of CsNO3 addition when most binding of 137Cs occurred was 700 mg equal to 0.0472 g 137Cs isotope, while the optimum weight addition of Lampung zeolite was 1000 mg until 1200 mg equal to 137Cs isotope content in137Cs-zeolite obtained was 0.0557 g. The 137Cs isotope separation process using both methods could be concluded that the cation exchange method using zeolite by the addition of 1000 mg Lampung zeolite until 1200 mg had better result than the precipitation method. The other advantages of this method were easier and safer than precipitation methods that the process should be in ice batch (-4oC), using chemicals HClO4, acetone and ethanol which had volatile and exothermic character. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8017 10.17146/urania.2016.22.2.3092 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 99-110 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8017/6141 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8019 2024-12-23T01:47:58Z urania:ART

PENGARUH PARAMETER PROSES ELEKTRODEPOSISI TERHADAP PENENTUAN BERAT ISOTOP 235U DALAM PEB U3Si2/Al PASCA IRADIASI Yanlinastuti, Yanlinastuti Boybul, Boybul Ginting, Aslina Br. Anggraini, Dian Electrodeposition 235U isotope Alpha spectrometer U3Si2/Al irradiated fuel plate Electrodeposition process for determination of 235U isotope is affected by current, contact time, feed volume and type of buffer solution. The aim of this research is to obtain optimum parameter of electrodeposition process for 235U determination in U3Si2/Al irradiated fuel plate. Electrodeposition process is done by some variation of parameters, that are the current 1; 1.2; and 1.4 ampere with variation of time 1; 1.5; 2.0; 2.5 and 3.0 hours. The standard solution volume of 20% U3O8 are 100; 300; 500; 700 µL with variation of electrolyte buffer solution NH4Cl, NH4OAc, NH4HSO4 dan (NH4)2SO4. The result of electrodeposition process showed that optimum condition electrodeposition process for 235U isotope was obtained currents 1.2 ampere with process time for 2 hours, and 300 µL volume feed with electrolyte buffer NH4HSO4 in pH 3,5. The results of the electrodeposition process was obtained 2.314 µg deposition of the isotope 235U with recovery of 99.91%. The condition optimum parameter above was used for electrodeposition process in determination of 235U isotope which contained in U3Si2/Al irradiated fuel element. The spectrometer alpha analysis was obtained weight of 235U isotope as 0.403 µg with recovery of 91.80%. The conclusion of this research, the electrodeposition methods is good enough to be used for determining 235U isotope in U3Si2/Al irradiated fuel plate. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8019 10.17146/urania.2016.22.2.3093 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 111-120 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8019/6143 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8020 2024-12-23T01:47:58Z urania:ART

PEMBUATAN RADIONUKLIDA MOLIBDENUM-99 (99Mo) HASIL AKTIVASI NEUTRON DARI MOLIBDENUM ALAM UNTUK MEMPEROLEH TEKNESIUM-99m (99mTc) Saptiama, Indra Herlina Sriyono Sarmini, Endang Abidin Kadarisman Molybdenum Technetium Radionuclide Separation Irradiation Uranium usage restriction causes the hospitals in Indonesia difficult to obtain the supply of 99mTc. At Present, 99mTc is obtained from molybdenum as a uranium fission product. Production of 99Mo radionuclide resulted from neutron activated natural molybdenum (MoO3) in G.A Siwabessy reactor could be used as an alternative method for producing 99mTc. The aim of this research is synthesized of   99Mo radionuclide from neutron activated natural molybdenum (MoO3) to obtain 99mTc. The five grams of MoO3 powder was packed in a quartz ampule and inserted into inner capsule then also inserted into outer capsule as a target material. It was irradiated in G.A Siwabessy reactor for 100hours. Based on theoritical calculation, about 65 % of maximum 99Mo activity could be recovered. After irradiation, MoO3 was dissolved by NaOH 4 M solution, so it was natrium molybdate (Na2MoO4) solution. 99Mo and 99mTc radionuclide were analyzed using gamma spectrometer. 99Mo radionuclide was detected on Na2MoO4 solution as product that had specific activity of 0.81 Ci 99Mo/ g Mo. 99mTc as daughter radionuclide was separated from 99Mo as parent radionuclide using separated column containing zirconium-based material (ZBM) as 99Mo adsorbent. 99mTc radionuclide has been successfully separated using ZBM column although recovery of 99mTc was quite low in which approximately 52 to 71 %. The 99mTc radionuclide was recovered in the form of sodium pertechnetate (NaTcO4) solution. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8020 10.17146/urania.2016.22.2.3094 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 121-132 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8020/6144 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8021 2024-12-23T01:48:41Z urania:ART

PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al Ajiriyanto, Maman Kartaman Ginting, Aslina Br. Supardjo Boybul, Boybul U−7Mo/Al Diffusion couple Interaction layer Microstructure DTA and microhardness U−7Mo/Al alloy had great potential as research reactor fuel, but it had several disadvantages, such as, it can form a interaction layer at the interface during the process of fabrication and irradiation in a reactor. The research objective was determining the interaction layer that was caused by interdiffusion or diffusion couple of U−7Mo with AlMg2 alloy which was annealed at 500 °C and 550 °C for 24 hours. The observation of microstructure used a Scanning Electron Microscope (SEM) on diffusion couple sample which was heated at temperature of 500 °C had not seen the layer interaction or the formation of a new phase between particles of U−Mo and Al matrix, but heating treatment at 550 °C indicated that U−7Mo alloy had been interdiffusion with AlMg2 plate produced (U,Mo) Alx compound on the interfaces. It was evidenced by interdiffusion reaction analysis used DTA that showed that U−7Mo / Al alloy at 500 °C had good heat compatibility, but at temperatures upper than 550 °C it had been phase changed from a + d to a + g phase. The heating in DTA furnace up to 679.14 °C produced U(Al,Mo)x meta stable phase and then interdiffusion process with uranium molten formed layer interaction that formed UAlx compound agglomerates (UAl4, UAl3 and UAl2). Agglomerates was formed from the heating process which was similar to agglomerates that caused by irradiation. U−7Mo / Al Fuel alloy that had 58% burn up had been interdiffusion between U−7Mo with Al matrix produced U(Al,Mo)x metastable phase that turned into (U, Mo) Al7 layer, UMo2Al20 precipitates, (UMo)Al3−Al and formed a boundary or UAlx (UAl4, UAl3 and UAl2) agglomerates. The results of microstructure analysis used SEM and interdiffusion reactions used DTA was supported by the analysis of micro hardness used Vickers Hardness. The results of hardness analysis that was done to AlMg cladding and U−7Mo alloy (before and after heating) and diffusion couple of U−7Mo / Al samples with AlMg2 plate after heating at 550 °C were respectively 64.62 and 340.45 HV (before heating) and 52.34; 303.16 and 497.34 HV (after heating). Diffusion couple U−7Mo/Al with AlMg2 plate samples had the highest hardness value. This hardness difference showed that the interdiffusion test used diffusion couple produced a new compound (U, Mo) Alx in interface zone that had different character, but the formation of interaction layer is not expected in the fuel U−Mo / Al dispersion because micro hardness and density of (U, Mo) Alx compound’s layer was lower than the average density of U−7Mo/ Al alloy. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8021 10.17146/urania.2016.22.1.2740 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-12 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8021/6145 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8022 2024-12-23T01:48:41Z urania:ART

PENGARUH UNSUR Zr TERHADAP PERUBAHAN SIFAT TERMAL BAHAN BAKAR DISPERSI U-7Mo-xZr/Al Supardjo Kadarjono, Agoeng Boybul, Boybul Ginting, Aslina Br. U-7Mo-xZr/Al dispersion fuel Phase transformation Entalphy Heat capacity Thermal properties data of nuclear fuel is required as input data to predict material properties change phenomenon during the fabrication process and irradiated in a nuclear reactor. Study the influence of Zr element in the U-7Mo-xZr/Al (x = 1%, 2% and 3%) fuel dispersion to changes in the thermal properties at various temperatures have been stiffened. Thermal analysis includes determining the melting temperature, enthalpy, and phase changes made using Differential Thermal Analysis (DTA) in the temperature range between 30 °C up to 1400 °C, while the heat capacity of U-7Mo-xZr alloy and U-7Mo-xZr/Al dispersion fuel using Differential Scanning Calorimeter (DSC) at room temperature up to 450 °C. Thermal analyst data DTA shows that Zr levels of all three compositions showed a similar phenomenon. At temperatures between 565.60 °C – 584.98 °C change becomes α + δ to α + γ phase and at 649.22 °C – 650.13 °C happen smelting Al matrix Occur followed by a reaction between Al matrix with U-7Mo-xZr on 670.38 °C – 673.38 °C form U (Al, Mo)x Zr. Furthermore a phase change α + β becomes β + γ Occurs at temperatures 762.08 °C – 776.33 °C and diffusion between the matrix by U-7Mo-xZr/Al on 853.55 °C – 875.20 °C. Every phenomenon that Occurs, entalphy posed a relative stable. Consolidation of uranium Occur in 1052.42 °C – 1104.99 °C and decomposition reaction of U (Al, Mo)x and U (Al, Zr)x becomes (UAl4, UAl3, UAl2), U-Mo, and UZr on 1328,34 °C - 1332,06 °C , The existence of Zr in U-Mo alloy increases the heat capacity of the U-7Mo-xZr/Al, dispersion fuel and the higher heat capacity of Zr levels increased due to interactions between the atoms of Zr with Al matrix so that the heat absorbed by the fuel increase. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8022 10.17146/urania.2016.22.1.2741 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 13-24 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8022/6146 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8023 2024-12-23T01:48:41Z urania:ART

PENGARUH PROSES SINTERING TERHADAP PERUBAHAN DENSITAS, KEKERASAN DAN MIKROSTRUKTUR PELET U-ZrHx Masrukan Mujinem Sintering Density Hardness Microstructure U-ZrHx pellet The process of sintering fuel pellets U-ZrHx performed to obtain a higher density compared pellets before sintered. At first, U-ZrHx pellets made from the content of Zr respectively 35%, 45% and 55% by weight. Sintering process is carried out for 1 hour at varying temperature 1100°C, 1200°C and 1300°C with an increase rate of 2°C temperature/ min, then lowered the temperature at a rate of decrease in temperature of 10°C / min until room temperature. U-ZrHx pellets that have been sintered subsequently subjected to various tests, such as: the dimensions, density, hardness, and microstructure. In testing the density, it appears that at the same Zr composition, but the sintering temperature is higher than an increase in the density. When seen in the same sintering temperature and Zr content is higher then the lower the density obtained. Micro hardness testing results show that the composition Zr 35% fixed and 55% were sintered at the sintering temperature is higher, have the higher the hardness, and then decreases as the sintering temperature increase. If the test results viewed in the sintering temperature fixed but the content of Zr is different, it can be seen from the content of Zr 35% (U-35ZrHx) to 45% (U-45ZrHx) an increase in hardness and will decrease to an increase in the content of Zr to 55% (U- 55ZrHx). Meanwhile, the results of the microstructure observation of the pellets can be seen that at composition Zr fixed but the sintering temperature is higher, then the distance between the powder / porosity increasingly dense. When viewed from the sintering temperature fixed but the composition increasingly higher Zr it is seen that higher Zr composition, the density between the powder in the pellets decreased. It can be concluded that the sintering process the pellets will increase density, hardness, and distance between powder increasingly dense. The optimum condition is achieved at sintering process the pellets with a composition of 45% Zr (U-45ZrHx) sintered at temperatures of 1200°C. In these conditions the sintered the pellets has a density of 8.673 g/cm3, hardness of 661 HVN without developing cracks. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8023 10.17146/urania.2016.22.1.2742 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 25-34 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8023/6147 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8024 2024-12-23T01:48:41Z urania:ART

PERBANDINGAN DENSITAS PELET UO2 HASIL PELETISASI MENGGUNAKAN SERBUK DAN MIKROSPIR Mutiara, Etty Rachmawati, Meniek Masrukan Pelletization UO2 microsphere Powder Density A pelletization process UO2 fuel has been developed using UO2 microsphere as a substitute of UO2 powder. Microspheres are spherical, free flowing and porous with certain hardness (soft particle). The benefit of using microsphere in pelletization process is dust free in compaction and more effective in packing so the granulation process and solid lubricants are not required. It is hypothesized that the use of UO2 microsphere in the pelletization process will provide higher sintered pellet density than UO2 powder at the same pelletization process parameters. UO2 microsphere size used in this pelletization was 900 μm with crushing strength of 2.0 N/particles while the UO2 powder size between 150 and 850 μm. The pelletization processes of UO2 microsphere and UO2 powder were performed by varying the compacting pressure between 200 MPa up to 500 MPa and sintered at temperatures of 1100 °C for 6 hours in an atmosphere of hydrogen and nitrogen gas mixture. Characterizations performed on the green and sintered pellets of UO2 microsphere and UO2 powder were dimension measurements, weighing and densities measurements. The densities of green and sintered pellets of UO2 microsphere were higher than the green and sintered pellets densities of UO2 powder with corresponded compaction pressure variations. The results indicate that the density of the green pellets both compaction results UO2 powder and UO2 microsphere increased with increasing compacting pressure. microsphere UO2 pellets density crude ranged from 82.1 to 84.2%TD. At the same sintering conditions, both Compaction UO2 powder and UO2 microsphere. Compaction shows the density increases with the greater pressure compacting process. The sintered pellets densities of UO2 obtained from this research were not appropriate with density requirements of PWR fuel. It is necessary to perform advanced research related to the effective pelletization process parameters and UO2 microsphere specifications in providing the appropriate sintered pellets densities. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8024 10.17146/urania.2016.22.1.2743 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 35-42 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8024/6148 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8025 2024-12-23T01:48:41Z urania:ART

PELAPISAN PERMUKAAN PELET UO2 DENGAN ZIRKONIUM DIBORIDA MENGGUNAKAN METODA SPUTTERING Sungkono Atmono, Tri Mardji Priyantoro, Dwi UO2 pellet ZrB2 layer Sputtering Microstructure Thickness Crystal structure Chemical composition Development of the nuclear fuel technology was aimed to improve nuclear power plant efficiency. One of the proposed solutions was utilization nuclear fuel with the high burn up. This leads to an increase in gas fission and nuclear reactor core reactivity. The excess reactivity of the reactor core controlled by integrated fuel burnable absorber. In connection it has made a layered thin UO2 pellets with burnable absorber. The objectives of this research was to obtain the characters of zirconium diboride layer on UO2 pellets surface i.e microstructure, crystal structure and chemical composition. The results of this experiment showed that the microstructure of UO2 + 0.4% Cr2O3 pellet has equiaxial and acicular grains with a diameter of 2.44 mm, whereas the UO2 + 0.3% Nb2O5 pellets has a grain structure in the form of equiaxial and flattened rod with a diameter of 2.47 mm. The layer of zirconium diboride on the surface of UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets are similar namely thin and compact with a thickness of 2.71 mm and 2.82 mm. The identification of X-ray diffraction pattern on UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets to exhibit there are of UO2 phase with cubical crystal structure and phase ZrB2 phase with hexagonal crystal structure. Meanwhile, zirconium concentration on the surface layer of UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets each of 1.82 mg and 1.90 mg, the presence of zirconium element show that ZrB2 layer formed on the surface of UO2 pellets. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8025 10.17146/urania.2016.22.1.2744 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 43-52 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8025/6149 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8026 2024-12-23T01:48:41Z urania:ART

ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI Kuntjoro, Sri Udiyani, Pande Made Inventory RDE 10 MWt power 20 MWt power 30 MWt power Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term, and it will be used as the basis for calculating the distribution of radionuclides into the environment. Badan Riset dan Inovasi Nasional 2024-12-10 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8026 10.17146/urania.2016.22.1.2745 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 53-64 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8026/6150 Copyright (c) 2016 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8674 2025-06-12T09:49:12Z urania:ART

ENHANCING CORROSION RESISTANCE OF AISI–1010 IN RSG–GAS SECONDARY COOLING PIPES THROUGH INHIBITOR COMPOUND INTERVENTION Nurlia, Enung Kusumastuti, Rahayu Aziz, Abdul Ajiriyanto, Maman Kartaman Kriswarini, Rosika Priyotomo, Gadang Nikitasari, Arini Sunaryo, Geni Rina Utomo, Setyo Budi Sriyono Supriyadi Prifiharni, Siska Musabikha, Siti secondary cooling pipes inhibitors corrosion behavior RSG-GAS The secondary cooling pipes of RSG-GAS utilize AISI-1010, facilitating water circulation directed to the cooling tower after heat absorption from the primary coolant in the heat exchanger. The subsequent dissipation of heat through the cooling tower into the ambient air aims to uphold the primary coolant's temperature below 40°C. This study focuses on evaluating the impact of inhibitor compounds on the corrosion behavior of AISI 1010. Employing FTIR and GCMS techniques, the inhibitor compounds were analyzed, complemented by visual and SEM examinations for surface morphology. The corrosion rate, influenced by the inhibitor, was quantified using a potentiostat. FTIR analysis revealed a spectrum of functional groups, encompassing O-H, N-H, aromatic compounds, C=O, C-O, and phosphate. Nitrogen, oxygen, and phosphorus elements within the inhibitor exhibited binding interactions with iron in the material. GCMS identified prominent compounds like 1H-Benzotriazole (RTs 14.4469 & 14.7746) - which is a major corrosion inhibitor indicated by its high peak and area percentage, Trans-1-methyl-2-nonyl-cyclohexane (RT 6.6834) as primary constituents of the inhibitor. Visual inspection post-immersion in a 150-ppm inhibitor solution showcased a lack of corrosion products on AISI 1010, contrasting with visible corrosion on the material without an inhibitor. SEM analysis confirmed a protective layer formation. The addition of the inhibitor at 150 ppm significantly enhanced AISI 1010's corrosion resistance, evidenced by a 46.71% reduction in the corrosion rate. This underscores the efficacy of the inhibitor in mitigating corrosion effects on AISI 1010, contributing valuable insights for materials in similar cooling system applications. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8674 10.55981/urania.2025.8674 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 13-24 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8674/8918 Copyright (c) 2025 Enung Nurlia, Rahayu Kusumastuti, Abdul Aziz, Maman Kartaman Ajiriyanto, Rosika Kriswarini, Gadang Priyotomo, Arini Nikitasari, Geni Rina Sunaryo, Setyo Budi Utomo, Sriyono, Supriyadi, Siska Prifiharni, Siti Musabikha https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8675 2025-07-02T06:28:27Z urania:ART

PERILAKU KOROSI BAJA KARBON RENDAH PIPA PENDINGIN SEKUNDER REAKTOR DALAM INHIBITOR ZINK FOSFAT Kriswarini, Rosika Khotib, Mohammad Insani, Andon Sutriah, Komar Noviarty Nugroho, Arif Nurlia, Enung Corrosion Carbon steel Inhibitor Carbon steel, one of the fuction in nuclear industry is used as a material for the cooling pipe structure of nuclear reactor operating systems. In use, the pipe is susceptible to corrosion due to contact with cooling water. Inhibitors of zinc phosphate solution derived from a mixture of zinc chloride and phosphoric acid with concentration variations of 25, 35, 50, 75 and 100 ppm in 0.1% NaCl are used as materials to study the effect of corrosion rates of carbon steel. Electrochemical methods are helpful for observing the corrosion behavior of carbon steel in zinc phosphate media. Methods include Open Circuit Potential (OCP) for 600 seconds, Electrochemical Impedance Spectroscopy (EIS) in the frequency range of 105 Hz to 0.01 Hz and Potentiodynamic Plot (PDP) in the voltage range of -1 Volt to 1 Volt using potentiostat. To support the results of corrosion tests, morphological observations were carried out using SEM, chemical composition of substrat using XRF and C/S Analyzer and zinc phosphate solution content using AAS, UV-Vis and FTIR. The results obtained show that at a concentration of 35 ppm, the corrosion rate gives a lower value than at other concentrations. The corrosion rate is 11.7 mpy Badan Riset dan Inovasi Nasional 2024-08-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8675 10.55981/Urania.2024.8675 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8675/6675 Copyright (c) 2024 Rosika Kriswarini, Mohammad Khotib, Andon Insani, Komar Sutriah, Noviarty, Arif Nugroho, Enung Nurlia https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8677 2025-07-02T06:32:46Z urania:ART

MODEL SIMULATION OF THE EFFECT OF ZINC PHOSPHATE INHIBITOR ON CORROSION RATE OF SECONDARY COOLING PIPE FOR PRESSURIZED WATER REACTOR USING PREDICTOR-CORRECTOR METHOD Al Afghani, Fajar Kartaman Ajiriyanto, Maman Kriswarini, Rosika Nugroho, Arif Carlos Sihotang, Juan Nurlaily, Ely Sari Dwi Putri, Anditania Model Simulation Corrosion Carbon steel Inhibitor ZnPO4 Corrosion is one of the unavoidable problems in nuclear reactor cooling systems. Secondary Cooling Pipe Multi-Purpose Reactor G.A Siwabessy (MPR-GAS) based on low carbon steel is very easy to oxidize because it is an open circulation cooling system. Zinc-phosphate inhibitor (ZnPO4) is one of the inhibitors used to inhibit the corrosion reaction that occurs between the low carbon steel cooling pipe and the cooling water environment. The research has been carried out on numerical calculations to find a solution for the corrosion rate model of steel with and without an inhibitor by the Predictor-corrector method. The simulation results show that the effect of ZnPO4 inhibitor on the corrosion rate of steel is seen as a relationship between inhibitor concentration and stable time, if the inhibitor concentration increases, the stable time will also increase. The rate of corrosion that occurs (reducing the concentration of corroded product). Therefore, the recommended inhibitor concentration value is 0.9% for models in 0.1 M HCL medium. This concentration is the most appropriate concentration to slow down the corrosion rate most minimal. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8677 10.55981/Urania.2024.8677 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-20 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8677/6676 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8678 2025-07-02T06:32:55Z urania:ART

PENGUKURAN BERAT RADIONUKLIDA 137Cs DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI DENGAN METODE KOLOM KROMATOGRAFI Boybul Yanlinastuti Noerpitasari, Erlina Carlos Sihotang, Juan Fauzi Rahmatullah, Helmi Nugroho, Arif Sukron Fajrin Husein, Mohammad Sigit, Rohmad Br Ginting, Aslina WIdodo, Susilo Irradiated U3Si2/Al 137Cs Cation exchange method Zeolite The burn up value of the fuel is one of the parameters that must be known in the post-irradiation examination. The determination of burn up aims to determine the performance of U3Si2/Al fuel after being used in the reactor. One of the isotops that can be used as an indicator of burn up is 137Cs because it has a long half-life (T1/2) of 30.17 years with a fission yield of 6.26%. Therefore, in this study, a separation was carried out to determine the amount of 137Cs in the U3Si2/Al fuel solution post irradiation using the cation exchange column method using zeolit from Lampung along with its recovery rate. The solution of irradiated U3Si2/Al fuel from Top (T), Middle (M), and Bottom (B) position, was each pipetted in a duplicate manner (duplo) in an amount of 50 µL and added with 2 mL of 0.1 M HNO3. The solutions were then measured activity using a gamma spectrometer to obtain the initial weight of 137Cs radionuclide. The solution was then inserted into a chromatography column containing 700 mg of zeolit Lampung and allowed to flow under vacuum condition. The separation results showed that 137Cs was bound to the zeolit in the form of 137Cs-zeolit solid phase, while other isotops remained in the liquid phase. The solid 137Cs-zeolit and 137Cs in the supernatant were then measured with a gamma spectrometer to determine the weight of 137Cs bound to the zeolit. The measurement results showed that the average initial weight of 137Cs (before separation) were 7.49 × 10-3 µg/gFEP and dan 6.97 × 10-3 µg/gFEP for T1 and T2 from top section, 10.50 × 10-3 µg/gFEP for M1 from Middle section, and dan 8.43 × 10-3 µg/gFEP and 8.76 × 10-3 µg/gFEP for B1 and B2 from Bottom section, respectively. After separation, the average weight of 137Cs were 7.43 × 10-3 µg/gFEP and 6.82 × 10-3 µg/gFEP for T1 and T2, 10.10 × 10-3 µg/gFEP for M1, and 8.32 × 10-3 µg/gFEP and 8.41 × 10-3 µg/gFEP for B1 and B2, respectively. From the weight of 137Cs before and after separation, the average separation recovery with the cation exchange chromatography column method using zeolit was obtained at 97.748%. This recovery value indicates that zeolit Lampung is highly selective for absorbing 137Cs in the solution of irradiated U3Si2/Al fuel. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8678 10.55981/urania.2024.8678 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 21-30 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8678/6677 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8679 2025-07-02T06:33:06Z urania:ART

OPTIMASI PEMISAHAN SESIUM DI DALAM PELAT ELEMEN BAKAR TIPE U3Si2/Al PASCA IRADIASI MENGGUNAKAN METODE PENGENDAPAN KLOROPLATINAT Nugroho, Arif Sutriah, Komar Mulijani, Sri Basuki, Triyono Ade Rofiqi, Ahmad Noviarty Fauzi Rahmatullah, Helmi Kurniati Suryaman, Ganisa Guswardani Chloroplatinate Separation Post-irradiation Cesium U3Si2/Al Several methods for separating fission product ¹³⁷Cs from uranium isotopes (²³⁸U, ²³⁵U, ²³⁶U, and ²³⁴U) in U₃Si₂/Al fuel have been widely developed. Optimization and separation were performed using the chloroplatinate precipitation to establish a new cesium (Cs) separation method. The method developed in this research offers the potential to achieve higher-purity cesium precipitates. This study aims to develop a method of separating the ¹³⁷Cs isotope from U₃Si₂/Al fuel element plates (PEB) using chloroplatinate reagents. The procedure involves dissolving the pre-irradiation U₃Si₂/Al PEB and adding several elements such as Sm, Sr, Cs, and Nd to represent the composition of pre-irradiation PEB, simulating its original condition. The cesium precipitation process in the supernatant solution was carried out by adding HCl and NaOH to separate impurities from cesium. The precipitation was conducted by adding H₂PtCl₆ to precipitate and isolate cesium from other elements. The precipitation process was optimized by varying the temperature and precipitation time. The optimum precipitation conditions were achieved at a temperature of -3°C and a time of 10 minutes, resulting in a Cs2PtCl6 precipitate weighing 0.3361 g. XRD and SEM analyses were used to complement and verify the cesium separation results using the chloroplatinate precipitation method, showing a uniform distribution of all elements. The purity of the Cs2PtCl6 precipitate obtained was 100%, with a diffractogram pattern identical to the reference. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8679 10.55981/urania.2024.8679 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 31-44 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8679/6678 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8680 2025-07-02T06:33:15Z urania:ART

PENENTUAN KEMURNIAN RADIONUKLIDA DARI PRODUK GENERATOR Mo-99/Tc-99m NON-FISI MENGGUNAKAN SPEKTROMETER GAMMA Fadila, Nurul Febriana, Siska Nur Pratiwi, Beta Molybdenum-99 Technetium-99m Mo-99/Tc-99m Generator Gamma Spectrometer The medical radionuclide technetium-99m (Tc-99m) is the type of radionuclide that is most widely used in diagnostic processes in the world of nuclear medicine. This is related to its characteristic of having a half-life of 6 hours and releasing a single gamma energy (E ) of 140 keV. Tc-99m is a product of the Mo-99/ Tc-99m generator and is a decay child of molybdenum-99 (Mo-99) which has a half-life of 66 hours. The Mo-99  radionuclide can be produced via a neutron activation reaction which requires the non-fission natural target molybdenum trioxide (MoO3) as a target in a nuclear reactor, so it is called a non-fission Mo-99/Tc-99m generator. The aim of this research is to determine the purity value of the Tc-99m product from the non-fission Mo-99/Tc-99m generator using the gamma spectrometer method. This is done by analyzing the energy and efficiency calibration of the gamma spectrometer using a liquid standard source (Liquid Source) of a mixture of Ba-133  and Eu-152, and a point standard source (Point Source) of Eu-152, according to the geometry for the measurements of Mo-99 breakthrough and Tc-99m products respectively for each three-generator elution process. The neutron activation process of natural molybdenum (MoO3) was carried out in the nuclear reactor G.A. Siwabessy, Indonesia. The purity of the Tc-99m  product radionuclide from the non-fission Mo-99/Tc-99m generator was obtained at 99.999% for each elution while the largest Mo-99 breakthrough was 0,0018 µCi Mo-99/mCi Tc-99m and the lowest was 1,04 x 10-4 µCi Mo-99/mCi Tc-99m. The Mo-99 breakthrough results obtained were below the value of 0,15 µCi Mo-99/mCi Tc-99m set by the US Pharmacopoeia, so the resulting product of Tc-99m was good, and showed that the product could be used for the labeling process to become a radiopharmaceutical. This radionuclide purity measurement method can be used to determine quality control parameters for non-fission Mo-99/Tc-99m generator products. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8680 10.55981/urania.2024.8680 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 45-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8680/6679 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/8681 2025-07-02T06:28:10Z urania:ART

ANALISIS PENGARUH GEOMETRI FUEL ASSEMBLY TERHADAP KRITIKALITAS DAN DISTRIBUSI FLUKS NEUTRON DALAM TERAS REAKTOR RSG-GAS MENGGUNAKAN OPENMC Fathan Rastri, Ibnu Octadamailah, Saga Sukarman RSG-GAS Criticality Flux distribution OpenMC Fuel element geometry The G. A. Siwabessy Multi-Purpose Reactor (RSG–GAS) is included in the MTR type research reactor having a straight plate-shaped fuel element fueled by uranium silicide with a density of 2.96 gU/cm3. This research aims to analyze the influence of the geometry of the straight plate model fuel element on the criticality and distribution of neutron flux in the RSG–GAS reactor core. Monte Carlo simulations with OpenMC were carried out on five types of fuel elements namely from RSG–GAS, HANARO, WWR-SM, MARIA, and BRR, with mass U235 250 grams per fuel element. The research results show that the geometric shapes of different fuel elements influence the criticality value and flux distribution in the same core. The highest criticality value was obtained for the fuel element from HANARO followed by RSG-GAS and the lowest was obtained for the fuel element from WWR-SM. The flux distribution in the fuel elements from MARIA and BRR has a better flux distribution. Badan Riset dan Inovasi Nasional 2024-12-23 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/8681 10.55981/urania.2024.8681 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-68 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/8681/6680 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9125 2025-06-12T09:53:31Z urania:ART

CORROSION CHARACTERISTICS STUDY ON URANIUM AND ITS ALLOYS (U-Nb, U-Zr, AND U-Zr-Nb) Masrukan Adi, Wisnu Ari Handoko, Djati Djuhana, Dede corrosion U Metal U-Zr alloy U-Nb U-Zr-Nb Uranium metal (U) and its alloys, such as U-Zr, U-Nb, and U-Zr-Nb, can be used as fuel in nuclear reactors, especially in research reactors. These fuels have various characteristics, including physical, chemical, mechanical, and corrosion properties. Among these characteristics, fuel corrosion properties play a critical role during the reactor operation, after use, and also during the storage period. Fuel damage due to corrosion is influenced by several factors, such as the concentration of oxidizing agents, the flow rate of corrosive fluids, temperature, and acidity (pH). The corrosion process in uranium metal produces hydrogen gas (H₂), which reacts with uranium to form pyrophoric uranium hydride (UH₃),  and uranium oxidation produces UO₂.  If the oxidation process continues for an extended period, the uranium metal will be depleted. The corrosion potential (Ecorr) and corrosion rate (CR) of uranium metal in a passive state do not depend on pH until the pH reaches 7. The corrosion potential increases once the pH exceeds 7 and continues to rise to pH 11. Uranium alloys with low Zr, Nb, or Ru content, when alloyed with Zr and Nb up to 2.5 wt.% and Ru up to 5.0 wt.%, show minimal impact on the corrosion potential (Ecorr) in the pH range of 4.0–9.0 unless there is a dissolved oxidant. U-Nb alloys with higher Nb content, exhibit greater resistance to corrosion and depleted uranium (without Nb). Under acidic conditions (pH 1.18), U-6Zr-2Nb alloys demonstrate the best corrosion resistance. In demineralized water, U-6Zr-5Nb alloys show the best corrosion resistance, while under basic conditions (pH 11.02), U-6Zr-8Nb alloys exhibit the best corrosion resistance. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9125 10.55981/urania.2025.9125 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 35-48 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/9125/8942 Copyright (c) 2025 Masrukan, Wisnu Ari Adi, Djati Handoko, Dede Djuhana https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9131 2025-07-02T06:43:52Z urania:ART

PENGARUH IRADIASI GAMMA TERHADAP KARAKTERISTIK POLIMER POLIESTER TAK JENUH UNTUK APLIKASI SOLIDIFIKASI LIMBAH RADIOAKTIF Setiawan, Risdiyana Sugita, Purwantiningsih Farid, Muhammad Ratiko Rachmadetin, Jaka Heriyanto, Kuat Purwanto, Yuli Saputra, Dwi Luhur Ibnu Setiawan, Andry Solidification Unsaturated polyester Irradiation The first effort in radioactive waste management is through radioactive waste characterization by identifying nuclides and measuring radioactivity. Characterization of radioactive waste is helpful in waste management consisting of waste containment, waste decomposition, temporary waste storage, and disposal operations. The use of polyester polymers in radioactive waste solidification is one of the important solutions in safe and sustainable waste management. These polymers are chosen for their good mechanical and chemical properties, such as resistance to degradation and long-term stability. This study aims to characterize and analyze the effect of gamma irradiation on the mechanical and chemical properties of unsaturated polyester polymers used in radioactive waste solidification. Irradiation was carried out with dose variations of 50, 100, 200, and 300 kGy. The results showed that at low to medium doses (50-200 kGy), there was an increase in compressive strength and density due to the formation of crosslinks between polymer chains. However, at higher irradiation doses (300 kGy), material degradation occurred leading to a decrease in strength. Raman spectrophotometry and SEM showed changes in chemical structure and morphology due to irradiation, especially in terms of changes in functional groups and structural density. The results of this study indicate that irradiation can strengthen polymers at certain doses, but dose limits must be considered to prevent material degradation. Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9131 10.55981/urania.2024.9131 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 69 – 80 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9131/6991 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9141 2025-07-09T02:36:34Z urania:ART

IMPROVEMENT OF PLASMA ELECTROLYTIC OXIDATION (PEO) COATING ON ZIRCONIUM ALLOYS USING ADDITIVES FOR NUCLEAR FUEL CLADDING APPLICATION Carlos Sihotang, Juan Kartaman Ajiriyanto, Maman Anawati Plasma Electrolytic Oxidation Zircaloy Additives Corrosion This study explores the improvement of Plasma Electrolytic Oxidation (PEO) coatings on zirconium alloys using additives for pressured water reactor (PWR) fuel cladding application. PEO is a promising surface treatment for zirconium alloys, offering improved corrosion and wear resistance, essential for applications in extreme environments, such as nuclear reactors. Various types of stable particle and dissolved additives were incorporated into the electrolyte during PEO process to modify the coating's microstructure, phase composition, and mechanical properties. The incorporation of stable particle such as Al₂O₃, CeO₂, and Y₂O₃ not only reduces surface porosity and roughness but also increases the presence of the tetragonal zirconia phase (t-ZrO₂), leading to enhanced corrosion and wear resistance. Additionally, the use of dissolved additives, including complex agents and salts, positively affected plasma characteristics and further improved coating density and durability. Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9141 10.55981/urania.2024.9141 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 81 – 96 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9141/6992 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9142 2025-07-02T06:43:52Z urania:ART

ANALISIS RADIONUKLIDA 137Cs DI DALAM PELAT ELEMEN BAKAR NUKLIR U3Si2/Al DENSITAS 4,8 gU/cm3 BURN UP 60% PASCA IRADIASI MENGGUNAKAN METODE KLOROPLATINAT Nugroho, Arif Sutriah, Komar Mulijani, Sri Basuki, Triyono Ade Rofiqi, Ahmad Boybul Noerpitasari, Erlina Kriswarini, Rosika Carlos Sihotang, Juan Jamaludin, Agus Chloroplatinate Gamma spectrometry SEM U3Si2/Al 137Cs isotope Analysis of 137C isotopes in post-irradiation nuclear fuel is vital in determining burn up and fuel behavior during fission. The 137Cs isotope is used as input data for burn up calculations because it has a long half-life (30.17 years) and a large fission yield (6.026%). This study uses the chloroplatinate precipitation method to separate the 137Cs isotope from U3Si2/Al nuclear fuel with a density of 4.8 gU/cm³ with a burn up of 60% after irradiation. The separation process begins with cutting and dissolving the fuel sample, followed by the addition of CsNO₃ as a carrier solution and H₂PtCl₆ reagent to form Cs₂PtCl₆ precipitate. The results of the analysis using gamma spectrometry showed that the spectrum of the Cs₂PtCl₆ precipitate had the highest peak area of ​​67.6298 cps compared to the brown residue of 0.9165 cps and the supernatant of 0.2450 ​​cps. In addition, the results of this study showed an increase in the separation recovery value from 70% to 96.39%, which indicates the success of the separation of the 137Cs isotope from the U3Si2/Al fuel solution with a density of 4.8 gU/cm3 using the chloropaltinate precipitation method. The results of elemental mapping of the Cs₂PtCl₆ precipitate using SEM (scanning electron microscope) showed an even distribution of platinum (Pt), cesium (Cs), and chlorine (Cl) components, which shows the efficiency of this precipitation method. In addition, EDS (energy dispersive X-ray spectroscopy) analysis shows the elemental composition of the Cs₂PtCl₆ deposits. The three main elements identified are Pt, Cl, and Cs. The results of this study can provide an essential contribution to evaluating the performance of U3Si2/Al fuel with a density of 4.8 gU/cm3 after irradiation. They can be used for further research on Cs isotopes. Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9142 10.55981/urania.2024.9142 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 97 – 108 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9142/6993 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9143 2025-07-02T06:43:52Z urania:ART

FORMULASI INHIBITOR KOROSI BERBASIS ZINK–FOSFAT DAN MIKO–DEA PADA PIPA PENDINGIN SEKUNDER REAKTOR SERBA GUNA GERRIT AUGUSTINUS SIWABESSY Kriswarini, Rosika Insani, Andon Khotib, Mohammad Sutriah, Komar Nugroho, Arif Carlos Sihotang, Juan Saputra, Ade Corrosion Carbon steel zinc-phosphate and miko-DEA inhibitors RSG-GAS The material for the secondary cooling pipes of RSG-GAS is low carbon steel. The pipes carry fluids that have the potential to cause corrosion. To reduce the corrosion rate, both inorganic and organic inhibitors are added. The inhibitors used are a mixture of zinc-phosphate as an inorganic inhibitor and miko-DEA as an organic inhibitor in a 0.1% NaCl medium. Zinc-phosphate inhibitors are tested at concentrations of 25, 35, and 50 ppm, and miko-DEA inhibitors at concentrations of 50, 75, and 100 ppm. To study the effects on corrosion rate and the corrosion behavior of carbon steel, electrochemical methods were employed, including Open Circuit Potential (OCP) for 600 seconds, Electrochemical Impedance Spectroscopy (EIS) over a frequency range from 105 Hz to 0.01 Hz, and Potentiodynamic Plot (PDP) over a voltage range from -1 Volt to 1 Volt using a potentiostat. To support the corrosion test results, surface morphology of the samples with and without inhibitors was observed using SEM, and compounds in the zinc-phosphate and Miko-DEA solutions were identified using FTIR. The optimal formulation that resulted in the lowest corrosion rate was a mixture of 50 ppm zinc-phosphate and 75 ppm miko-DEA in a 0.1% NaCl medium. The achieved corrosion rate was 10.57 mpy. Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9143 10.55981/urania.2024.9143 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 109 – 122 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9143/6994 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9144 2025-07-02T06:43:52Z urania:ART

PENGOLAHAN LIMBAH RADIOAKTIF CAIR AKTIVITAS TINGGI MELALUI VITRIFIKASI GELAS BOROSILIKAT Wati Rachmadetin, Jaka Purwanto, Yuli Saputra, Dwi Luhur Ibnu Ratiko Aisyah Artiani, Pungky Ayu Vitrification High activity liquid radioactive waste Borosilicate glass High-level liquid radioactive  waste in Indonesia is presently generated from the Mo99 radioisotope production process. Until present, the treatment of the waste has not been carried out and it is still stored in the hot cell awaiting for the establishment of the methods, technology and processing facilities. Several countries have implemented a vitrification process to process LRCAT. Several types of glass have been developed for LRCAT vitrification, namely alumina silicate glass, phosphate glass, borosilicate glass and ceramic glass. Borosilicate glass is the most ideal type of glass to use. This is because the process of making borosilicate glass is simpler and the results of vitrification of borosilicate waste-glass have chemical resistance and physical resistance as well as high waste content. The objective of this research is to determine the composition of borosilicate glass-forming materials through a vitrification process suitable for imobilation of high-level waste from 99Mo production that meets the requirements of the final waste form. Several requirements in choosing a matrix for LRCAT vitrification are: easy and practical manufacturing process, high waste content, high chemical resistance, high radiation stability, high heat stability, good physical integrity and long-term stability. The vitrification process uses borosilicate glass forming materials with two different compositions. The vitrification process was carried out at a temperature of 1150 ºC for 2.5 hours then cooled to room temperature. Glass characterization is carried out by leaching test and X-ray diffractometer (XRD). Waste glass with GF-1 glass forming material with various waste contents cannot be melted at a temperature of 1150 ºC. The vitrification process was unsuccessful and no melting of the waste-glass occurred. The glass forming material GF-1 cannot be used for LRCAT vitrification and no further research was carried out. This is related to its use on an industrial scale, if the melting temperature is higher than 1150 ºC, then the refractory corrosion rate is very high so the melter life is shorter. This will generate more solid radioactive waste (secondary waste). In addition, high melting temperature will produce more gas so that gas handling is more complex. Meanwhile, glass material with GF-2 glass forming material can be melted at a temperature of 1150 ºC. GF-1 waste-glass were not leached test because the vitrification process was unsuccessful and no melting of the waste-glass occurred. The leaching test results of all GF-2 glass-waste compositions that were made ranged from 0.79 x 10-4 gram.cm-2.day-1 to 2.25 x 10-4 gram.cm-2.day-1. The leach rate requirement for PNC Japan standard borosilicate glass is 2.30 x 10-5 gram.cm-2.day-1 and based on Pacific Northwest National Laboratory (PNNL) is 10-4 – 10-7 gram.cm-2.day-1. Based on the results of chemical resistance characterization (leachate test) and physical integrity, the composition of GF-2 glass-waste with a waste content of 20% by weight provides the most optimal characteristics Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9144 10.55981/urania.2024.9144 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 123 – 138 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9144/6995 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9148 2025-07-02T06:43:52Z urania:ART

DETEKSI CACAT PADA OBJEK LOGAM DI INDUSTRI NUKLIR MENGGUNAKAN MACHINE LEARNING Nugroho, Tri Sulistiyo Hari Effendy, Nazrul Kusnanto Objects Metal Nuclear industry Defect detection Machine learning Objects made from metal are quite widely used in the industrial world, both nuclear and non-nuclear related. In the manufacturing process of this object, there is possibility of production defects occurring. Apart from that, metal objects can also experience defects during the operation of the object, for example due to impact, fracture, or due to corrosion or other chemical reactions. In the nuclear industry, metal objects include reactor tank, control rod, pump, pipes, and nuclear reactor fuel cladding. In order to ensure the continued optimal functionality of equipment, machinery, and/or installations constructed from metal, it is essential to implement a comprehensive maintenance programme. Such a programme must include regular inspections to identify any potential damage that may affect the integrity of the metal components. One method of detecting defects in technical objects that is currently being developed is using artificial intelligence or machine learning, either combined with visual inspection and analysis methods, X-ray, gamma ray, electromagnetic or ultrasonic results. Machine learning itself has been applied to several applications. The research carried out in this article used bibliographic analysis and was visualized using VOSviewer. This paper presents methodologies for the non-destructive analysis of metal objects, with a specific focus on visual analysis and machine learning. The integration of test data and visual analysis with machine learning enables the formulation of informed conclusions regarding the condition of metal objects, thereby facilitating decision-making processes. Badan Riset dan Inovasi Nasional 2024-12-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9148 10.55981/urania.2024.9148 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 139 – 152 2528-0473 0852-4777 10.55981/urania.v30i2 eng https://ejournal.brin.go.id/urania/article/view/9148/6996 Copyright (c) 2024 Author(s) https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9199 2025-12-08T07:54:47Z urania:ART

The Effect of Current Density on The Surface Morphology in the Thin Coating Process of Nickel on Zr-2 Using the Electroplating Method Wicaksono, Yusuf Gigih Ridwan Azwar Manaf electroplating nickel zirconium-2 thin film current density    After the Fukushima-Daiichi reactor accident in 2011, one of the research and development focuses of nuclear fuel worldwide has been on coatings for Enhanced Accident Tolerant Fuels (EATF). Chromium (Cr) coatings are considered suitable due to their high oxidation resistance; however, Cr has limitations, particularly its poor diffusion on certain materials such as zirconium (Zr). Nickel coatings are therefore used as an interlayer to overcome the diffusion problem of chromium on zirconium substrates. Several surface coating methods are available, such as physical vapor deposition (PVD), chemical vapor deposition (CVD), high velocity oxygen fuel (HVOF), detonation gun (D-Gun), and electroplating. Electroplating was chosen in this study because of its high productivity, simple equipment, and low cost. The purpose of this research was to investigate the effect of current density on the surface morphology in the electroplating of Zr-2 with nickel. Electroplating is a process in which metal ions in an electrolyte solution are driven by an electric field to deposit onto a material. The electrolyte solution used in this study consisted of 200 g/L NiSO₄·6H₂O as the nickel source, 35 g/L NiCl₂·6H₂O as the activator, and 30 g/L H₃BO₃ as the pH buffer. The current densities applied were 0.015 A/cm², 0.025 A/cm², and 0.05 A/cm². After deposition, hardness tests were conducted, and the surface morphology was examined using SEM and EDS. The results showed that increasing the current density led to larger average nickel grain sizes, namely 4.68 μm, 6.19 μm, and 6.84 μm, as well as larger pore areas on the surface, namely 6.5 μm², 20.85 μm², and 27.98 μm². Micro-Vickers hardness tests indicated that higher current density increased hardness values, measured at 163.48 HV, 178 HV, and 234.25 HV, respectively. Cross-sectional SEM analysis revealed that the coating produced at a current density of 0.015 A/cm² showed better quality compared to higher current densities. This was evidenced by smaller pore areas, thinner coating thickness (7.44 μm compared to 19.17 μm and 8.42 μm), and the absence of defects at the coating–substrate interface, which were observed at 0.025 and 0.05 A/cm². To achieve thickness values closer to calculations, uniform spacing between the cathode and anode as well as the use of a fresh electrolyte solution are recommended. The use of a nickel interlayer can be a promising option to improve the surface performance of Zr-2.  Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9199 10.55981/urania.2025.9199 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 83-90 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/9199/11051 Copyright (c) 2025 Yusuf Gigih Wicaksono, Ridwan, Azwar Manaf https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9210 2026-10-08T02:11:44Z urania:ART

Ligand Interaction in Thorium Adsorption by Pacitan Bentonite Laksmana, Roza Indra Khotib, Mohammad Mas'ud, Zainal Alim Mellawati, June Pratama, Afiq Azfar Indryati, Suci Saputra, Dwi Luhur Ibnu thorium adsorption acid activation ligands radioactive waste bentonite pacitan Many studies have investigated the use of bentonite as an adsorbent; however, questions remain regarding the effects of sulfate and phosphate on thorium (Th) adsorption. Sulfate may reduce adsorption through the formation of Th-sulfate complexes and competition for adsorption sites between SO2-and UO2+, whereas phosphate is an important component of subsurface systems and plays a key role in controlling Thmobility. This study investigated Thadsorption by natural and acid-activated Pacitan bentonite as adsorbents for radioactive waste treatment. Because of its large surface area and porosity, bentonite has a high adsorption capacity, which can be further enhanced by acid activation. The adsorbents were characterized by FTIR and XRD to evaluate structuralchanges after activation. Batch experiments were conducted to examine the effects of pH, contact time, temperature, and complexing ligands, including sulfate, carbonate, and phosphate, on Thadsorption. The results showed that natural bentonite exhibited high Thadsorption efficiency from the beginning of the experiment, whereas acid-activated bentonite showed a gradual increase in adsorption efficiency over time. Thcomplexation with sulfate decreased adsorption efficiency, while phosphate improved Thstability in activated bentonite. These findings indicate that Pacitan bentonite, in both natural and acid-activated forms, is a promising adsorbent for Th-containing radioactive waste, provided that the adsorption conditions are properly optimized. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9210 10.55981/urania.2026.9210 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-10 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/9210/12249 Copyright (c) 2026 Roza Indra Laksmana, Mohammad Khotib, Zainal Alim Mas'ud, Afiq Azfar Pratama, Suci Indryati, June Mellawati, Dwi Luhur Ibnu Saputra https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9213 2025-12-08T08:37:16Z urania:ART

Uranium Adsorption Behavior on Pacitan Bentonite Laksmana, Roza Indra Khotib, Mohammad Mas'ud, Zainal Alim Mellawati, June Pratama, Afiq Azfar Indryati, Suci Saputra, Dwi Luhur Ibnu uranium adsorption acid activation bentonite ion exchange nuclear waste Uranium adsorption by Pacitan bentonite was investigated to evaluate its capacity for capturing uranium ions under various conditions. Before being used as an adsorbent, Pacitan bentonite was characterized in terms of pore structure, surface area, surface morphology, adsorption capacity, and chemical composition. This study aims to analyze the factors that influencing uranium adsorption on Pacitan bentonite, including contact time, pH, temperature, and the presence of ligands in solution. Both natural and acid-activated Pacitan bentonite were examined to compare their adsorption performance. The results showed that Pacitan bentonite effectively adsorbs uranium, with the highest adsorption occurring under acidic to neutral pH conditions and decreasing at higher pH levels. Acid activation reduced the uranium adsorption capacity due to structural and surface property changes. In addition, the addition of ligands such as NaCl, Na2SO4, and Na2HPO4 influenced uranium adsorption, with Na2SO4 having the most pronounced effect compared to NaCl and Na2HPO4. Overall, the results indicate that Pacitan bentonite has strong potential as a uranium adsorbent material, especially under suitable conditions to maintain ion exchange and surface adsorption. These findings provide important insights for developing adsorbent materials for nuclear waste treatment and remediation of uranium-contaminated environments.  Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9213 10.55981/urania.2025.9213 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 91-102 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/9213/11052 Copyright (c) 2025 Roza Indra laksmana, Mohammad Khotib, Zainal Alim Mas'ud, June Mellawati, Afiq Azfar Pratama, Suci Indryati, Dwi Luhur Ibnu Saputra https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/9350 2025-06-13T12:51:05Z urania:ART

INTEGRATION OF SAFETY ANALYSIS USING HAZOPS AND FTA ON PRECIPITATION PROCESS IN URANIUM REFINING AND CONVERSION FACILITY Oktavianto, Putra Saputra, Ade Sholikhah, Mu'nisatun Uranium refining and conversion facility Precipitation HIRADC HAZOPS FTA Implementation of safety at nuclear facilities must be implemented as well as possible. This is related to safety, especially for operators who carry out the process directly. In the uranium refining and conversion facility, safety analysis using HIRADC has been implemented to support safety in carrying out the process, especially in the uranyl nitrate precipitation process. However, this is not enough to ensure that safety during the precipitation process at the uranium refining and conversion facility can be maintained. To complement the existing HIRADC, another safety analysis is needed. Therefore, HAZOPS and FTA were analysed. This HAZOPS analysis will complement the safety side of the facility using an indicator control system that will automatically eliminate the risk of hazards that occur during the operation of the facility, while the FTA will analyse the causes of failure that can cause accidents so that the causal factors of the accident can be identified. From the results of the HAZOPS analysis, it was found that there was 1 potential hazard that fell into the low risk category, and 11 potential hazards that fell into the medium risk category. While the results of the FTA analysis obtained 15 events that cause product failure in the precipitation tank. From the results of the safety analysis that has been carried out, it can be concluded that HAZOPS and FTA can complement HIRADC to ensure safety can be maintained during the process. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/9350 10.55981/urania.2025.9350 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 57-66 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/9350/8913 Copyright (c) 2025 Putra Oktavianto, Ade Saputra, Mu'nisatun Sholikhah https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/10793 2025-06-12T09:54:03Z urania:ART

ANALYSIS OF DENSITY, PHASE, AND MICROSTRUCTURE IN U–Mo ALLOYS Octadamailah, Saga Suryaman, Ganisa Kurniati Supardjo, Supardjo Mustika, Deni Sihotang, Juan Carlos Gigih Wicaksono, Yusuf Pribadi, Slamet Boybul Isnaini, Anissa Langenati, Ratih Sutarya, Dede U-Mo neutron flux phase research reactor microstructure The performance of research reactors heavily depends on the intensity of the neutron flux produced, which is a key factor in reactor efficiency and effectiveness. To enhance the performance of research reactors, high-density fuel is required to increase neutron flux in the reactor core, enabling optimal irradiation processes. One of the most promising fuel candidates is uranium-molybdenum (U-Mo) alloy. This study aims to investigate the properties of U-Mo alloys through a series of characterizations, including density measurements, phase analysis using X-ray Diffraction (XRD), and microstructural observations using a Scanning Electron Microscope (SEM). The density measurements revealed values close to theoretical calculations, with densities of 17.423 g/cm³ for U-7Mo, 17.292 g/cm³ for U-8Mo, and 17.229 g/cm³ for U-9Mo. XRD analysis identified the presence of the γ-U phase with a body-centered cubic (bcc) crystal structure, characterized by diffraction peaks at 2θ angles of 37°, 53°, 66°, and 78°. SEM observations showed a pattern of bright and dark areas, which were further analyzed using Energy Dispersive Spectroscopy (EDS). EDS analysis indicated that the bright areas had a higher concentration of Mo and a lower oxide content compared to the dark areas. These findings suggest that the addition of Mo not only improves the structural homogeneity but also enhances the oxidation resistance of U-Mo alloys. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/10793 10.55981/urania.2025.10793 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 49-56 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/10793/8915 Copyright (c) 2025 Saga Octadamailah, Ganisa Kurniati Suryaman, Supardjo Supardjo, Deni Mustika, Juan Carlos Sihotang, Yusuf Gigih Wicaksono, Slamet Pribadi, Boybul, Anissa Isnaini, Ratih Langenati, Dede Sutarya https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/11218 2025-06-12T09:52:02Z urania:ART

INFLUENCE OF Mo AND Nb CONTENT ON THE MECHANICAL AND CORROSION PROPERTIES OF AlCrMoNbZr HIGH-ENTROPY ALLOYS FOR NUCLEAR FUEL CLADDING Hardiyanti, Heri Naibaho, Marzuki Kurniawan, Budhy Setiawan, Jan Ginting, Masno Insani, Andon Widakdo, Januar High Entropy Alloys (HEAs) AlCrMoNbZr crystal structure microstructure hardness corrosion The effect of varying Mo and Nb content in AlCrMo5%NbZr and AlCrMoNb5%Zr on their mechanical and corrosion properties has been investigated. X-ray diffraction (XRD) analysis revealed the presence of cubic and hexagonal crystal structure, with AlCrMoNb5%Zr exhibiting a higher concentration of hexagonal structures, contributing to improved hardness and corrosion resistance. Scanning electron microscopy (SEM) and energy-dispersive spectroscopy (EDS) confirmed a dendritic morphology with uniform elemental distribution for both samples. Vickers hardness and the corrosion testing shows the AlCrMoNb5%Zr has superior hardness and significantly lower corrosion rate compared to AlCrMo5%NbZr, respectively. These findings suggest that less amount of Nb content enhances mechanical and corrosion properties, making AlCrMoNb5%Zr a potential material for nuclear reactor applications, mainly as fuel cladding material. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/11218 10.55981/urania.2025.11218 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 25-34 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/11218/8917 Copyright (c) 2025 Heri Hardiyanti, Marzuki Naibaho, Budhy Kurniawan, Jan Setiawan, Masno Ginting, Andon Insani, Januar Widakdo https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/11422 2026-09-28T03:13:58Z urania:ART

DEVELOPMENT OF MATERIALS FOR IMMOBILIZATION OF CESIUM AND CERIUM BASED ON A COMBINATION OF NOCHAR SUPER ADSORBENT AND UNSATURATED POLYESTER FOR PRODUCTION WASTE RADIOISOTOPE 99Mo Setiawan, Risdiyana Sugita, Purwantiningsih Farid, Muhammad Ratiko Aisyah Rachmadetin, Jaka Heriyanto, Kuat Purwanto, Yuli Saputra, Dwi Luhur Ibnu Artiani, Pungky Ayu Unsaturated polyester Nochar waste loading This research aims to develop unsaturated polyester-based materials combined with Nochar superabsorbent as radioactive waste immobilization materials. The focus is to lock up radioactive elements such as cesium (Cs) and cerium (Ce), which are used as surrogates for 99Moradioisotope waste. This study evaluates the performance of immobilized materials in adsorbing and stabilizing radioactive elements using various waste loading percentages of 10%, 15%, 20%, and 25%. The characterization results showed that the combination of unsaturated polyester and Nochar was able to absorb the waste effectively without any migration of radioactive elements to the environment, according to the leaching test results. However, based on the compressive strength test, a high increase in waste loading resulted in a decrease in the mechanical strength of the composite. Thus, an optimal waste loading of 20% provides a balance between absorption capacity and mechanical strength. This study demonstrates the potential of using unsaturated polyester and Nochar as safe and effective immobilization materials for radioactive waste, especially in the context of long-term radioactive waste management. Badan Riset dan Inovasi Nasional 2025-05-31 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/11422 10.55981/urania.2025.11422 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 1-12 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/11422/8916 Copyright (c) 2025 Risdiyana Setiawan, Purwantiningsih Sugita, Muhammad Farid, Ratiko, Aisyah, Jaka Rachmadetin, Kuat Heriyanto, Yuli Purwanto, Dwi Luhur Ibnu Saputra, Pungky Ayu Artiani https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/11503 2025-12-08T09:24:38Z urania:ART

Resrad Parameter Sensitivity Simulation To Ascertain Local Parameters for Safety Assesment on Bangka Island Setiawan, Andry Kurniati, Mersi Iskandar, Dadong Pratama, Hendra Adhi Purwanto, Yuli Putra, Zico Pratama radiation dose safety assessment RESRAD parameter sensitivity TENORM Tin slag and waste generated by tin processing often contain Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM), which pose significant radiation hazards to workers, the public, and the natural ecosystem. To mitigate these risks, the development of a dedicated landfill facility is crucial for managing the radiological impact. This study analyzes the implications of storing TENORM waste, both prior to and following the construction of a landfill. The challenge of TENORM management persists, primarily due to the presence of long-lived radionuclides include uranium-238 (U-238), thorium-232 (Th-232), and potassium-40 (K-40), which are often inadequately handled by industry. To address this, a landfill design was proposed and evaluated by software simulation using RESRAD-Offsite 4.0 software. The default parameter values are multiplied by 0.25, 0.5, 1, 1.5, and 1.75. The resulting dose calculations are compared to those using the default value. If the difference is below 10%, it is considered low sensitivity, between 10% and 1000%, medium sensitivity, and above 1000%, high sensitivity. The reserch aim to optimize safety and resource allocation,. From 194 parameters were simulated, with approximately half exerting a significant influence on dose estimations. This study used software simulations to determine the level of influence each parameter has on dose calculations. Parameters who have low sensitivity are soil to plant transfer factor for Th and U, radionuclide spesific release for Th, U, and K, medium sensitivity are precipitation, total porosity saturated zones, high sensitivity are Fraction of time spent on primary contamination at indoor and fraction of time spent for livestock.  Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/11503 10.55981/urania.2025.11503 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 113-121 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/11503/11054 Copyright (c) 2025 Andry Setiawan, Mersi Kurniati, Dadong Iskandar, Hendra Adhi Pratama, Yuli Purwanto, Zico Pratama Putra https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/12906 2025-12-08T07:48:48Z urania:ART

A Review on High-Temperature Steam Oxidation Resistance of Zirconium Cladding Materials Subjected to Plasma Electrolysis Nashrah, Nisa Marbun, Cahya Krisnawati Saputra, Akbar Ridho Sihotang, Juan Carlos Pressurized Water Reactor cladding materials zirconium alloys plasma electrolysis steam oxidation resistance  Zirconium alloys, such as Zircaloy-4 and ZIRLO, are the standard cladding materials in pressurized water reactors (PWR) due to their low neutron absorption and corrosion resistance. Yet, under loss-of-coolant accident conditions, rapid steam oxidation above 900 °C accelerates hydrogen uptake, embrittlement, and cladding failure. Plasma electrolysis (PE) has emerged as a promising surface modification strategy, directly converting the Zr surface into a ZrO₂-based ceramic layer with strong adhesion, phase stability, and enhanced oxidation resistance. This review provides a comprehensive overview of PE coating formation, emphasizing anodic oxidation, plasma microdischarge, and incorporation of electrolyte-derived elements that tailor microstructure and tetragonal ZrO₂ stabilization. Comparative assessments show that PE-coated claddings delay breakaway oxidation, suppress oxygen diffusion, and maintain structural integrity better than bare alloys or many physical vapor–deposited coatings. The influence of coolant chemistry, irradiation effects, and thermal cycling on long-term coating durability is also evaluated. Remaining challenges include controlling thickness, mitigating phase transformation, and ensuring irradiation stability. Addressing these issues will be critical to realizing PE-coated zirconium as a viable accident-tolerant fuel cladding for advanced nuclear reactors.  Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/12906 10.55981/urania.2025.12906 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 67-82 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/12906/11050 Copyright (c) 2025 Nisa Nashrah, Cahya Krisnawati Marbun, Akbar Ridho Saputra, Juan Carlos Sihotang https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/13399 2025-12-08T09:25:26Z urania:ART

A Critical Review of Radioactive Substances Security Management in Indonesia: Regulatory Framework and Implementation Challenges Based on Actual Incidents Wirasapta, Andicho Haryus Pamungkas, Tiara Deta radioactive substances security ionizing radiation safety government regulation The utilization of radioactive sources in various sectors, including medicine and industry, offers significant benefits to public welfare but also poses serious risks if not managed securely. This review article highlights the urgent need for a comprehensive understanding of radioactive substances security, which is conceptually distinct from radiation safety. The purpose of this review is to critically analyze the change of Indonesia's regulatory framework for radioactive substances security, evaluate the implementation challenges, and formulate policy recommendations based on domestic and international case studies. The methodology employed is a systematic literature review and comparative case study analysis of scientific literature, recent regulations, and incident reports from credible institutions such as the Nuclear Energy Regulatory Agency of Indonesia (BAPETEN) and the International Atomic Energy Agency (IAEA). The analysis aims to identify trends, gaps, and causal relationships between policy and practices. It had been found that regulatory framework in Indonesia, particularly through Government Regulation (GR) No. 45 Year 2023 and BAPETEN Regulation (BR) No. 5 Year 2024, has evolved to specifically address security threats. However, implementation faces significant challenges such as insufficient number of human resources and large geographic area. Case studies of the Goiânia incident and the Serpong incident demonstrate that failures in security management can have fatal and widespread consequences. It is concluded that strengthening radioactive substances security management in Indonesia requires not only a robust regulatory framework but also strategic investment in technological innovation, enhancement of qualification and competency personnel, and the establishment of a solid security culture. Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/13399 10.55981/urania.2025.13399 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 103-112 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/13399/11053 Copyright (c) 2025 Andicho Haryus Wirasapta, Tiara Deta Pamungkas https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/13522 2025-12-08T09:19:47Z urania:ART

Effect of Gamma Irradiation on Crosslink Density and Chemical Characteristics of Gadolinium Oxide-Natural Rubber Vulcanizate Composite Amanda, Dira Kemala, Tetty Rachmadetin, Jaka Giat Sukaryo, Sulistioso Cifriadi, Adi Khotib, Mohammad Heriyanto, Kuat Ramadhani, Achmad FTIR gamma irradiation Gd₂O₃ natural rubber SiO2 Natural rubber is a polymer that has been widely used in various industrial applications because of its high flexibility. In this study, vulcanizates were developed from SiO2-reinforced natural rubber with added Gd₂O₃. Vulcanizate composites containing 0, 5, and 10 phr of Gd₂O₃ were fabricated to evaluate the combined effect of inorganic fillers and gamma irradiation. All composites were irradiate with gamma rays at a dose of 100 kGy and were evaluated for their chemical characteristic, swelling behavior, and crosslink density. Fourier Transform Infrared (FTIR) analysis showed that gamma irradiation induces significant structural changes in the natural rubber vulcanizate. These changes were confirmed by the emergence of new absorption peaks assigned to O-H and C=C groups in the FTIR spectra. Swelling and crosslink density measurements showed that gamma irradiation increased the vulcanizate’s crosslink density, indicating the formation of a stiffer polymer network. On average, the crosslink density increased by approximately 11%, accompanied by a 9.07% decrease in swelling. These finding highlight the synergistic role of SiO2 and Gd₂O₃ under gamma irradiation in improving the integrity of natural rubber based composites.  Badan Riset dan Inovasi Nasional 2025-11-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/13522 10.55981/urania.2025.13522 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; 122-128 2528-0473 0852-4777 eng https://ejournal.brin.go.id/urania/article/view/13522/11055 Copyright (c) 2025 Dira Amanda, Tetty Kemala, Jaka Rachmadetin, Sulistioso Giat Sukaryo, Adi Cifriadi, Mohammad Khotib, Kuat Heriyanto, Achmad Ramadhani https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/13606 2026-10-08T02:11:44Z urania:ART

Determination of the Coral Age from Keraton Buton Fort, Baubau City, Based on 14C Activities Using Liquid Scintillation Counting Method Maming Zakir, Muhammad Tenrisa'na, Andi Rais, Zulfhadyanti Ramadani, Sitti Syara Sapar, Ajuk 14C specific activities carbon dating liquid scintillation counting coral Keraton Buton Fort Determination of the age of the corals that make up Keraton Buton Fort based on 14C activity using the LSC (Liquid Scintillation Counting) method has been carried out. The coral samples analyzed were the corals that made up the fort taken from the gate of Keraton Buton Fort and a coral sample from the beach of Makassar Island Baubau, asmodern coral. Sample preparation was carried out physically and chemically using a mixture of NaOH with 30% H2O2, followed by a mixture of HClO4with 30% H2O2, and finally with 5% HCl solution to produce a clean and whiter-looking sample with a weight reduction of 10.5% -15.8%. The carbonate of coral samples was separated as CO2by reaction with 10% HCl and absorption by KOH 1 N solution as K2CO3. The total carbon in the two fort coral sample and modern coral was 0.129, 0.132, and 0.129 grams, respectively, and was obtained through the titration method. The specific activity of 14C in the samples was 13.95 DPM/gC and 14.38 DPM/gC in both fort coral samples, and 15.48DPM/gC in modern coral. The age of the coral samples of Keraton Buton Fort calculated fromthe 14C specific activity data was between 817 ± 0.123 and 565 ± 0.105 years, which is older than the ages of the Keraton Buton Fort according to the historical reviews. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/13606 10.55981/urania.2026.13606 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 23-28 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/13606/12251 Copyright (c) 2026 Maming, Muhammad Zakir, Andi Tenrisa'na, Zulfhadyanti Rais, Sitti Syara Ramadani, Ajuk Sapar https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/13640 2026-10-08T02:11:44Z urania:ART

Transient Activation Encapsulation Enables Separation-Free Production of High-Purity 177Lu in the G.A. Siwabessy Medium-Flux Research Reactor Maskur, Maskur Sugiharto, Yono Sarmini, Endang Sekaringtyas, Fransiska Christydira Sulaiman Billah, Aulia Arivin Pujianti, Lira Aprilia Triyatna, Fani Ritawidya, Rien Ridwan, Muhammad Lutetium-177 direct production encapsulation radionuclidic purity separation-free The production of high-purity ¹⁷⁷Lu via the direct (n,γ) route is highly dependent not only on isotopic enrichment of the target material but also on the strategic selection of encapsulation materials to reduce radionuclidic impurities. In order to optimize the production of high-specific-activity ¹⁷⁷LuCl₃, this work combines a double-layer containment system consisting of quartz ampoules (SiO₂) and high-purity aluminum capsules (Al 1070) with an enriched Lu₂O₃target (74% ¹⁷⁶Lu). Irradiation was performed in the RSG-GA Siwabessy reactor (BRIN, Indonesia) at a thermal neutron flux of 2 × 10¹⁴ n·cm⁻²·s⁻¹ for 99 and 193 hours. The target was dissolved in HCl/H₂O₂to form ¹⁷⁷LuCl₃and then characterized for radionuclidic purity using high-resolution gamma spectrometry (HPGe) at 48 hours post-end of irradiation (EOI). The results exhibit that this integrated system produces products without detectable container-derived radionuclidic impurities at 48 h post-EOI: activation products from the quartz (¹⁹O, T½ = 26.47s; ³¹Si, T½ = 2.62 h) and aluminum (²⁸Al, T½ = 2.24 min) containers decayed completely within 24 hours. Under normal HPGe conditions, the metastable isomer ¹⁷⁷ᵐLu (T½ = 160.4 d, σ = 2.85 barn), the sole co-produced radionuclidic species, is present at <1% of total activity at EOI, is not detected spectroscopically from initial state of ¹⁷⁷Lu, which can be clinically negligible during the typical therapeutic window (1–14 days post-production). The specific activity obtained was 11 718 mCi/mg (CV = 5.9%) after 193 hours of irradiation, a level adequate for effective radiolabeling of monoclonal antibodies and peptides in targeted radionuclide therapy, despite the presence of stable Lu carrier. This study demonstrates that the interaction between target chemistry and container physics, rather than irradiation parameters alone, is a critical determinant in producing pharmaceutical-grade ¹⁷⁷Lu that meets international radionuclidic purity standards (>99.9%). This approach provides a scalable, separation-free production model applicable to medium-flux research reactors globally. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/13640 10.55981/urania.2026.13640 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 11-22 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/13640/12250 Copyright (c) 2026 Maskur Maskur, Yono Sugiharto, Endang Sarmini, Fransiska Christydira Sekaringtyas, Sulaiman, Aulia Arivin Billah, Lira Aprilia Pujianti, Fani Triyatna, Rien Ritawidya, Muhammad Ridwan https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/13659 2026-10-22T04:29:05Z urania:ART

Bioaccumulation of Zinc by Mantis Shrimp (Harpiosquilla Raphidea) in Seawater: The Application of Nuclear Technique Using Radioactive Metal Chairuman Budiawan Suseno, Heny Pujiyanto, Anung bioaccumulation radiotracer 65Zn mantis shrimp Harpiosquilla raphidea Zinc (Zn) is a heavy metal that, in certain amounts, can be toxic to aquatic organisms and humans. Mantis shrimp (Harpiosquilla raphidea) is a source of seafood for humans, but can accumulate heavy metals. This study investigated the ability of Harpiosquilla raphidea (H. raphidea) to accumulate Zn metal at various concentrations, salinity levels, and weights (wet basis) using 65Zn as a radioactive tracer. Radioactive tracer techniques offer quicker and more accurate results while minimizing the use of organisms. The tested Zn concentrations ranged from 0.5 to 2.0 ppm, salinity varied from 25 to 31 ppt, and weights ranged from 15 to 40 grams. The values ​​observed were uptake constant (ku), elimination constant (ke), bioconcentration factor (BCF) and biological half-life (tb1/2). In experiments with concentration variations, increasing concentration increased ku, ke, and BCF, but decreased tb1/2. While in experiments with salinity variations, increasing salinity decreased ku, ke, and BCF, but increased tb1/2 values. Experiments with variations in H. raphidea weight, increasing body weight decreased ku and ke, but increased BCF and tb1/2. In short, changes in Zn concentration, salinity, and biota weight affect Zn bioaccumulation in H. raphidea. Badan Riset dan Inovasi Nasional 2026-10-22 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/13659 10.55981/urania.2026.13659 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 2 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 61-73 2528-0473 0852-4777 10.55981/urania.v32i2 eng https://ejournal.brin.go.id/urania/article/view/13659/13246 Copyright (c) 2026 Chairuman, Budiawan, Heny Suseno, Anung Pujiyanto https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/14937 2026-10-08T02:11:44Z urania:ART

Comparative Study of Radiation Hardness in Semiconductor Materials (Si, SiC, GaN, Diamond) for Americium-241 Waste-Based Nuclear Batteries Adhi, Pribadi Mumpuni Mustaqima, Millaty Alpha-voltaic Americium-241 nuclear battery SRIM Using Americium‑241 (Am‑241) radioactive waste from spent smoke detectors as an alpha‑voltaic energy source provides a long‑term power solution for sensors in extreme environments. However, high‑energy alpha particles (5.48 MeV) are inherently destructive to semiconductor crystal structures. This study aims to identify the optimal Wide Bandgap (WBG) material by analyzing energy‑loss mechanisms (stopping power), lattice damage, and device geometry design parameters. SRIM 2013 simulations were performed on Si, SiC, GaN, and diamond targets. The analysis includesion distribution range,electronic and nuclear stopping profiles, the calculation of a Radiation Tolerance Index, and the determination of minimum active thickness (Tmin) based on projected‑range statistics. The results show that diamond exhibits the highest tolerance index of 670, enablingthe thinnest device design (~15 μm). In contrast, silicon requires nearly twice the active thickness (~28 μm) and suffers the most severe structural damage. Silicon carbide (SiC) is recommended as a cost‑effective alternative with a Radiation Tolerance Indexof 593. While diamond is the superior material for space‑constrained applications, SiC offers the best balance between radiation resistance and cost for terrestrial applications. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/14937 10.55981/urania.2026.14937 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 52-60 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/14937/12254 Copyright (c) 2026 Pribadi Mumpuni Adhi, Millaty Mustaqima https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/15001 2026-10-08T02:11:44Z urania:ART

The Effect of Time in the Anodizing Process on the Coating Characteristics and Corrosion Behavior of Zirconium Metal Sianturi, Manogari Al Afghani, Fajar Batubara, Frisca Ronauli Rahmadani, Sri Saputra, Romi zirconium anodizing corrosion resistance cladding PWR surface modification Zirconium and its alloys are the standard material for nuclear fuel cladding in Pressurized Water Reactors (PWR) due to their low neutron absorption cross-section, excellent mechanical properties, and good corrosion resistance in high-temperature water. However, the operational environment of a PWR imposes severe conditions that can degrade the cladding integrity over time, including oxidation, hydriding, and mechanical damage such as scratching or denting during fuel refueling operations. These surface defects can act as initiation sites for localized corrosion, potentially compromising the primary containment barrier. This study investigates the effectiveness of electrochemical anodizing as a surface modification technique to enhance the performance of Zirconium. The anodizing process was conducted at a constant voltage of 30 V with varying durations of 10, 15, and 20 minutes. The resulting surface characteristics were evaluated using Optical Microscopy, Digital Microscopy for roughness analysis, and X-Ray Diffraction (XRD). Mechanical reliability was assessed via Vickers Microhardness testing, while Corrosion behavior was studied in a 3.5% NaCl solution using Open Circuit Potential (OCP), Potentiodynamic Polarization (PDP), and Electrochemical Impedance Spectroscopy (EIS). The results demonstrated that increasing the anodizing time significantly improved the surface quality, reducing the arithmetic mean roughness Ra from 0.53 μm (10 min) to 0.24μm (20 min). XRD analysis confirmed the formation of a crystalline ZrO2oxide layer. Electrochemical tests revealed a substantial enhancement in corrosion resistance; the corrosion current density icorrdecreased by two orders of magnitude from 12.93 x 10-9A/cm2for the substrate to 0.19 x 10-9A/cm2for the 20-minute anodized specimen. The study concludes that a 20-minute anodizing treatment at 30 V produces a robust, smooth, and highly corrosion-resistant oxide layer suitable for mitigating degradation in nuclear fuel cladding applications. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/15001 10.55981/urania.2026.15001 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 41-51 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/15001/12253 Copyright (c) 2026 Manogari Sianturi, Fajar Al Afghani, Frisca Ronauli Batubara, Sri Rahmadani, Romi Saputra https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/15044 2026-10-08T02:11:44Z urania:ART

The Role of Aging-Workforce Management in Maintaining Safety Performance in High-Risk Nuclear Installations Putrajaya, Fadly Hendrian Utomo, Kabul Wahyu Meutia Aging workforce management, Safety behavior, High-risk installations, Nuclear safety The nuclear industry, as a high-reliability organization, faces serious challenges due to an aging workforce that could affect operational sustainability and safety performance. Although the literature on aging workforce management (AWM) is growing rapidly, empirical evidence linking AWM to safety performance in high-risk nuclear installations, particularly in developing countries, remains limited. This study aims to analyze the role of AWM in maintaining and improving safety behavior in nuclear installations in Indonesia. This study uses a quantitative, cross-sectional survey design with 281 employees from nuclear installations under the auspices of the National Research and Innovation Agency (BRIN). AWM is conceptualized as a multidimensional construct encompassing employee well-being and human resource practices, while safety behavior is measured through the dimensions of safety awareness and safety actions. Data were analyzed using Confirmatory Factor Analysis (CFA) and Structural Equation Modeling (SEM). The analysis results show that AWM has a positive and significant effect on safety behavior (β = 0.607; C.R. = 6.875; p < 0.001), with the measurement and structural models having a perfect fit. These findings confirm that well-being-oriented aging workforce management practices, continuous training, work ergonomics, and career planning can strengthen workers' safety awareness and compliance. Theoretically, this study extends the safety and human factors literature by positioning AWM as a strategic mechanism in maintaining the safety performance of high-risk socio-technical systems. In practice, the study's results provide policy implications for adaptive, sustainable human resource management to ensure the long-term safety of the nuclear industry in developing countries. Badan Riset dan Inovasi Nasional 2026-04-30 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/15044 10.55981/urania.2026.15044 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 29-40 2528-0473 0852-4777 10.55981/urania.v32i1 eng https://ejournal.brin.go.id/urania/article/view/15044/12252 Copyright (c) 2026 Fadly Putrajaya, Hendrian, Kabul Wahyu Utomo, Meutia https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/15550 2026-10-22T04:29:05Z urania:ART

Remediation of Radioactive Contamination in Soils Using Soil Washing Method: A Systematic Literature Review of Artificial and Real Contaminated Samples (2010-2025) Novalina, Ayusta Dea Mochamad Adhiraga Pratama Herry Wijayanto Soil washing removal efficiency radioactive contamination remediation PRISMA 2020 systematic review Soil contamination by radionuclides such as cesium (Cs) and uranium (U) poses significant environmental and health risks due to their long half-lives, approximately 30 years for Cs-137 and up to 4.5 billion years for U-238, enabling long-term persistence in soils. Soil washing is considered a promising remediation method due to its efficiency and rapid application. Although soil washing has been studied using both artificial and real contamianted soils, it remains unclear how well laboratory-prepared contamination represents contamination behavior in real soil. This uncertainty limits the use of laboratory removal efficiencies to predict soil-washing performance at real contaminated sites. This study aims to compare the efficiency of soil washing between artificial and real contaminated soils for Cs and U removal. A systematic literature review based on PRISMA 2020 guidelines was conducted on studies published from 2010 to 2025, with data analyzed using a thematic approach focusing on sample type, radionuclide characteristics, washing agents, and operational parameters. Of 54 records initially identified from ScienDirect (n=36) and Google Scholar (n=18), 15 articles met all inclusion criteria. The findings indicate that removal efficiency varies widely depending on soil type, radionuclide characteristics, initial concentration, pH, temperature, contact time, washing cycles, and washing agents. For cesium, artificial soils (n = 8 observations) yielded a mean removal of ~63%, whereas real soils (n = 6) yielded a lower mean of ~57%. For uranium, the only artificial soil study (n =1) showed a mean removal of ~75%, whereas real soils exhibited a mean removal of ~98%. Cesium is more difficult to remove due to strong adsorption to clay minerals, while uranium is more readily mobilized under acidic conditions. A key limitation of this review is that it was restricted to English-language and freely accessible full-text articles; therefore, the findings should be interpreted within the scope of the available open-access literature. This review highlights the need for direct comparison between artificial and real contaminated soils under the same experimental conditions, followed by field-scale validation before laboratory soil-washing results are generalized to real contaminated sites. Badan Riset dan Inovasi Nasional 2026-10-22 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/15550 10.55981/urania.2026.15550 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 2 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 74-93 2528-0473 0852-4777 10.55981/urania.v32i2 eng https://ejournal.brin.go.id/urania/article/view/15550/13247 Copyright (c) 2026 Ayusta Dea Novalina, Mochamad Adhiraga Pratama, Herry Wijayanto https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/16013 2026-10-22T04:29:05Z urania:ART

Effect of Ni Addition on the Density, Microstructure, Hardness, and Phase Stability of U–7Mo–xNi Alloys Diinar Hasnaa' Iftikhoor Ratih Langenati Ganisa Kurniati Suryaman Saga Octadamailah Muhammad Alfarisie Juan Carlos Sihotang Slamet Pribadi Mujinem Brigitta Verren Kasih Pramesti Nowo Riveli density microstructure hardness phase U-Mo-xNi alloy The Reduced Enrichment Research and Test Reactor (RERTR) program required a transition from high enriched uranium (HEU) to low enriched uranium (LEU). The most promising nuclear fuel with LEU is uranium-molybdenum (U-Mo). U-Mo has high density and can act as g-U stabilizer, but under irradiation test, showed undesirable interaction with aluminum (Al) matrix. To stabilizing g and minimize interaction with Al, Ni was added to U-7Mo. The study began with fabrication of U-7Mo-xNi alloy using an electric arc melting furnace under an argon atmosphere.  The resulting U-7Mo-xNi were characterize through microstructural observation using optical microscope, density measurement using automatic gas pycnometer, hardness testing using Vickers, and phase identification using x-ray diffraction (XRD) then compared to U-7Mo. The results show that U-7Mo-xNi has lower density than U-7Mo, with 17.42 for U-7Mo, 16.77 g/cm3 for U-7Mo-1Ni, 16.45 g/cm3 for U-7Mo-2Ni and 16.42 g/cm3 for U-7Mo-3Ni. The average grain size also shows a decrease from 20.97 µm to 14.41 µm, which leads to an increase in hardness based on Hall-Petch relations. The results of hardness test show the higher value than U-7Mo with value 240.4 HV, 313.87 HV, 340.51 HV and 357.47 HV for U-7Mo-xNi as Ni composition in order. For XRD analysis, peaks detected at 2θ angle of 37.0°, 53.6° and 66.5° same as U-7Mo peaks and identified as g-U phase, but in U-7Mo-xNi has a minor peak at 2θ angle of 40.8° identified as Mo. The addition of Ni to U-7Mo improve the hardness of the alloy and the g-U phase still be maintained. The dendritic structure and minor peaks in U-7Mo-xNi indicate the need for further homogenization and characterization. Badan Riset dan Inovasi Nasional 2026-10-22 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/16013 10.55981/urania.2026.16013 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 2 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 94-100 2528-0473 0852-4777 10.55981/urania.v32i2 eng https://ejournal.brin.go.id/urania/article/view/16013/13248 Copyright (c) 2026 Diinar Hasnaa' Iftikhoor, Ratih Langenati, Ganisa Kurniati Suryaman, Saga Octadamailah, Muhammad Alfarisie, Juan Carlos Sihotang, Slamet Pribadi, Mujinem, Brigitta Verren Kasih Pramesti, Nowo Riveli https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/16104 2026-10-22T04:29:05Z urania:ART

Microstructural Evolution and Corrosion Behavior of W-Coated Zr-4 Alloy Deposited by DC Sputtering at Different Voltages Jamaludin, Agus Wahyuni, Wulan Tri Bakhri, Syaiful Wicaksono, Yusuf Gigih Ajiriyanto, Maman Kartaman Sudjadi, Usman Setiati Achmadi, Suminar DC sputtering tungsten coating Zircaloy-4 corrosion nuclear cladding Zirconium (Zr) alloys are widely used as nuclear fuel cladding materials due to their low neutron absorption, high mechanical properties, and good corrosion resistance in high temperature aqueous environments. However, their performance might degrade under severe accident conditions. Tungsten (W), a refractory metal with high thermal stability and irradiation resistance, is a promising candidate material for protective coating. Nevertheless, the tungsten oxide film is not as stable in aqueous environments as the passive ZrO2 film. The effect of DC sputtering voltage on the microstructure, mechanical properties and corrosion behavior of W-coated Zr-4 substrates for nuclear applications was studied. To investigate the effect of deposition energy on coating characteristics, W coatings were deposited at 4, 5 and 6 kV. XRF and SEM-EDS analysis confirmed successful W deposition, with the sample at 5 kV having the most uniform coating coverage. SEM observations and cross-sectional analyses show that the coating microstructure and thickness are strongly dependent on the sputtering voltage. The coating thickness increased from ~4.8 to 7.8 µm at 4 and 5 kV and decreased to 5.2 µm at 6 kV. The coating density increased with increasing sputtering voltage, and at 6 kV the higher ion energy caused partial resputtering and, hence, a reduction in coating thickness. XRD results revealed the formation of nanocrystalline BCC (body-centered cubic) tungsten with crystallite sizes decreasing from 8.98 to 5.56 nm with increasing voltage. The hardness of Zr-4 substrate was only 187.63 ± 5.06 HV, while the coated sample reached 480-620 HV due to grain refinement and densification. The potentiodynamic polarization tests showed that the corrosion potential (Ecorr) shifted from −3.05 × 10⁻¹ V for the uncoated substrate to −3.12 × 10⁻¹ V for the 5 kV-coated sample, while the corrosion current density (Icorr) increased from 1.44 × 10⁻⁷ A/cm² to 4.86 × 10⁻⁷ A/cm². The results show lower corrosion resistance after coating deposition under the conditions explored. Nevertheless, the 5 kV coating showed the best compromise between mechanical enhancement and corrosion performance due to its compact microstructure and high hardness, suggesting the trade-off between the hardness enhancement and corrosion behavior. The increase in the corrosion current density is likely due to coating defects and the lower stability of tungsten oxide films compared with the protective ZrO₂ layer on uncoated Zr-4. Badan Riset dan Inovasi Nasional 2026-10-22 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/16104 10.55981/urania.2026.16104 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 2 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 101-113 2528-0473 0852-4777 10.55981/urania.v32i2 eng https://ejournal.brin.go.id/urania/article/view/16104/13249 Copyright (c) 2026 Suminar Setiati Achmadi, Agus Jamaludin, Wulan Tri Wahyuni, Syaiful Bakhri, Yusuf Gigih Wicaksono, Maman Kartaman Ajiriyanto, Usman Sudjadi https://creativecommons.org/licenses/by-sa/4.0

oai:ejournal.brin.go.id:article/16340 2026-10-22T04:29:05Z urania:ART

Bibliometric Mapping of Studies on Neutron Absorber Materials and Criticality Safety in Spent Nuclear Fuel Management Heriyanto, Kuat Manaf, Azwar Dwi Yudanto, Sigit Rachmadetin, Jaka Amanda, Dira bibliometric analysis spent nuclear fuel neutron absorber materials criticality safety dry storage Spent nuclear fuel storage systems must remove decay heat, maintain radiation shielding and confinement, and prevent nuclear criticality during interim storage, transportation, and disposal-oriented handling. Neutron absorber materials are central, maintaining subcritical margins in storage racks, casks, baskets, and transport packages. Although bibliometric studies examined nuclear fuel cycle technologies or spent nuclear fuel research, none comprehensively integrated spent nuclear fuel storage, neutron absorber materials, decay heat, and criticality safety, limiting systematic understanding of their research landscape and evolving relationships. This study maps Scopus-indexed research from 2000 to January 2026 on spent nuclear fuel storage, neutron absorber materials, and criticality safety. Scopus yielded 599 records on 27 January 2026; document-type and language screening retained 236 English-language journal articles. VOSviewer and Microsoft Excel analyzed annual publication trends, country and author productivity, journal distribution, keyword co-occurrence, thematic clusters, and temporal trends. Publications generally increased after 2014, peaking in 2021 and 2024. The United States, South Korea, and China led contributions; main venues included Annals of Nuclear Energy, Nuclear Technology, and Nuclear Engineering and Design. Three themes dominated: spent nuclear fuel storage technologies, neutron absorber and shielding materials, and computational analysis for nuclear criticality safety. Temporal trends shifted from reactor-related computational studies toward storage technologies and recent performance evaluation of advanced neutron absorber and shielding materials. This comprehensive bibliometric mapping reveals major themes, leading sources, productive authors, and emerging trends. Findings highlight the need to integrate material characterization, neutron testing, degradation assessment, and criticality modeling for long-term spent nuclear fuel storage safety and reliability. Badan Riset dan Inovasi Nasional 2026-10-22 info:eu-repo/semantics/article info:eu-repo/semantics/publishedVersion application/pdf https://ejournal.brin.go.id/urania/article/view/16340 10.55981/urania.2026.16340 Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir; Vol. 32 No. 2 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir; 114-128 2528-0473 0852-4777 10.55981/urania.v32i2 eng https://ejournal.brin.go.id/urania/article/view/16340/13250 Copyright (c) 2026 Kuat Heriyanto, Azwar Manaf, Sigit Dwi Yudanto, Jaka Rachmadetin, Dira Amanda https://creativecommons.org/licenses/by-sa/4.0