Transient Analysis of Simultaneous LOFA and RIA in RSG-GAS Reactor after 32 Years Operation

Authors

  • Muhammad Darwis Isnaini Center for Nuclear Reactor Technology and Safety, BATAN
  • Iman Kuntoro Center for Nuclear Reactor Technology and Safety, BATAN
  • Muhammad Subekti Center for Nuclear Reactor Technology and Safety, BATAN

DOI:

https://doi.org/10.17146/tdm.2020.22.3.5944

Keywords:

RSG-GAS, Simultaneous, LOFA, RIA, PARET

Abstract

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip. From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.

References

Anonymous, Safety Analysis Report of RSG-GAS Reactor, Rev. 10, National Atomic Energy Agency; 2008. (in Indonesian)

Hamid B.N., Hossen Md.A., Islam S.M.T., Begum R., Modelling an Unprotected Loss of Flow Accident in Research Reactor using Eureka-2/RR Code. J. of Physical Science. 2015. 26(2):73-87.

Pinem S., Sembiring T.M., Liem P.H., Neutronic and Thermal-Hydraulic Safety Analysis for the Optimization of the Uranium Foil Taget in the RSG-GAS reactor. Atom Indonesia. 2016. 42(3): 123-128.

https://doi.org/10.17146/aij.2016.532

Hakim A., Thermal-hydraulics analysis of Loss of Flow Accident of Multipurpose Reactor RSG-GAS with U3Si2Al Fuel. Prosiding Pertemuan dan Presentasi Ilmiah. Yogyakarta. 2000. 151-156. (in Indonesian)

Hastuti E.P., Sembiring T.M., Taryo T., Neutronic and Thermal Hydraulics Safety Analysis of RSG-GAS Core Conversion from Oxide to Silicide. Proc. Asian Physics Symposium. 2005. 335-341.

Abdelrazek I.D., Aly M.N., Badawi A.A., Elnour A.G.A., Benchmarking RSG-GAS reactor thermal hydraulic data using RELAP5 code. Ann. of Nucl. Energy. 2014. 70:36-43.

https://doi.org/10.1016/j.anucene.2014.02.023

Ekariansyah A.S., Hastuti E.P., Sudarmono., RELAP5 Simulation for severe accident analysis of RSG-GAS reactor. Tri Dasa Mega. 2018. 20(1): 23-34.

https://doi.org/10.17146/tdm.2018.20.1.4040

Hammoud A., Meftah B., Azzoune M., Radji L., Zouhire B., Amina M., Thermal-hydraulics behavior of the NUR Nuclear Research Reactor During a Fast Loss of Flow Transienst. J. of Nucl. Science and Tech. 2014. 51(9); 1154-1160.

https://doi.org/10.1080/00223131.2014.916235

Ibrahim S.M.A, El-Morshedy S.E.D., Abdelmaksoud A., Thermal hydraulics analysis of core flow bypass in a typical reasearch reactor. Nucl. Eng. and Tech. 2019. 51: 54-59.

https://doi.org/10.1016/j.net.2018.08.021

Talebi S., Najafi P., A two-phase model for simulation of MTR type research reactor during protected and unprotected LOFA. Progr. in Nucl. Energy. 2019. 100:274-288.

https://doi.org/10.1016/j.pnucene.2018.10.004

Dibyo S., Sudjatmi K.S., Sihana, Irianto I.D., Simulation of Modified TRIGA-2000 with Plate-Type Fuel under LOFA using EUREKA-2/RR Code. Atom Indonesia. 2018. 44(1):31-36.

https://doi.org/10.17146/aij.2018.541

Park C., Tanimoto M., Imaizumi T., Miyauchi M., Ito M., Kaminaga M., Preliminary accident analysis for a conceptual design of a 10MW Multi-purpose research reactor. JAEA-Technology 2012-039. Oarai. Januari 2013.

Hastuti E.P., Tukiran S., Widodo S., Sudarmono, Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes. Kerntechnik. 2018. 83(2): 96-105.

https://doi.org/10.3139/124.110844

Surbakti T., Pinem S., Suparlina L., Dynamic analysis on the safety criteria of the conceptual core design in MTR-type research reactor. Atom Indonesia. 2018. 44(2): 89-97.

https://doi.org/10.17146/aij.2018.545

Umbehaun, P.E., Torres, W.M., Souza, J.A.B., Yamaguchi, M., e Silva, A.T., de Mesquita, R.N., Scuro, N.L. and de Andrade, D.A. (2018) Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification. World Journal of Nuclear Science and Technology. 8: 54-69.

https://doi.org/10.4236/wjnst.2018.82006

Sudarmono, Hastuti E.P., Characterization of oxide fuel element temperature of RSG-GAS by using forced and natural convection cooling mode. Int. J. of Eng. And Science. 2018. 7(5): 49-56.

Guo Y.C., Wang G., Qian D., Yu H., Hu B., Guo S., Mi X.M., Ma J., Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT. Ann. of Nucl. Energy. 2018. 122: 125-136.

https://doi.org/10.1016/j.anucene.2018.08.031

Yari M., Lashkari A., Masoudi S.F., Hosseinipanah M., Three dimensional analysis of temperature effect on control rod worth in TRR. Nucl. Eng. and Tech. 2018. 50. 1266-1276.

https://doi.org/10.1016/j.net.2018.07.020

Hedayat A., Simulation and transien analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code. Nucl. Eng. and Tech. 2017. 49: 953-967.

https://doi.org/10.1016/j.net.2017.03.009

Foad B., Abdel-Latif S.H., Takeda T., Reactivity feedback effect on loss of flow accident in PWR. Nucl. Eng. and Tech. 2018. 50: 1277-1288.

https://doi.org/10.1016/j.net.2018.07.012

Darwis Isnaini M., Subekti M., Validation of SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition. Tri Dasa Mega. 2016. 18(1): 41-50.

https://doi.org/10.17146/tdm.2016.18.1.2367

INVAP SE, MTR_PC v3.0.A. Neutronic, thermal-hydraulic and shielding calculations of MTR-type reactors on personal computers. San Carlos de Bariloche: Nuclear Engineering Division, Argentina. 2006.

Tiyapun K., Wetchagarun S., Neutronic and thermal hydraulics analysis of TRIGA Mark II reactor using MCNP and COOLOD-N2 computer code. IOP Conf. Series: J. of Physics: Conf. Series 860 (2017) 012035.

https://doi.org/10.1088/1742-6596/860/1/012035

Hastuti E.P., Alfa S.K., Sudarmono, Analysis on the performance of the Bandung conversion fuel-plate TRIGA reactor in steady state with constant flow rate. Tri Dasa Mega. 2020. 22(2): 41-48.

https://doi.org/10.17146/tdm.2020.22.2.5843

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Published

2020-09-30

How to Cite

Isnaini, M. D., Kuntoro, I., & Subekti, M. (2020). Transient Analysis of Simultaneous LOFA and RIA in RSG-GAS Reactor after 32 Years Operation. Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega, 22(3), 111–119. https://doi.org/10.17146/tdm.2020.22.3.5944