VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION
DOI:
https://doi.org/10.17146/tdm.2016.18.1.2367Keywords:
transien, termohidraulika, PWR, simulator, COBRA-EN, MATPROAbstract
The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR) using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO) for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.
References
Subekti M., Darwis Isnaini M., Endiah P.H. The determination of radial fuel temperature in PWR-NPP simulator on steady state. Proceeding of Basic Research on Nuclear Science and Technology, PTAPB-BATAN, Yogyakarta;2012.p.76-81.
Darwis Isnaini M., Dibyo S., Suroso, Geni RS., Endiah P.H., Subekti M. Evaluation of core and subchannel thermal-hydraulics design parameter of AP1000 nuclear power plants on steady state condition. Journal of Reactor Technology of Tri Dasa Mega 2012; 14(1):14-28.
Anonymous The westinghouse pressurized water reactor nuclear power plant, Westinghouse Electric Corporation Water Reactor Divisions, Chapter 16: Plant Operation, (1984). p. 207- 218.
Darwis Isnaini M. The influence of nozzle and spacer grid against thermal-hydraulics parameters of AP1000 reactor fuel assemblies. Journal of Reactor Technology of Tri Dasa Mega 2013; 15(3): 159-170.
Darwis Isnaini M., Widodo S., Subekti M. The thermal-hydraulics analysis on radial and axial power fluctuation for AP1000 reactor. Journal of Reactor Technology of Tri Dasa Mega 2015; 17(2): 79-86.
https://doi.org/10.17146/tdm.2015.17.2.2290
Aghaie M., Zolfaghari A., Minuchehr M., Norouzi A. Enhancement of COBRA-EN capability for VVER reactor calculations. Annals of Nuclear Energy 2012; 46: 236-243.
https://doi.org/10.1016/j.anucene.2012.03.026
Rahmani Y., Pazirandeh A., Ghofrani M.B., Sadighi M. Calculation of the deterministic optimum loading pattern of the BUSHEHR VVER-1000 reactor using the weighting factor method. Annals of Nuclear Energy 2012; 49: 170-181.
https://doi.org/10.1016/j.anucene.2012.05.023
Zarifi E., Jahanfarnia G., Veysi F. Thermal-hydraulics modeling of nanofluids as the coolant in VVER-1000 reactor core by the porous media approach. Annals of Nuclear Energy 2013; 51: 203-212.
https://doi.org/10.1016/j.anucene.2012.07.041
Zarifi E., Jahanfarnia G., Veysi F. Subchannel analysis of nanofluids application to VVER 1000 reactor. Chemical Engineering Research and Design 2013; 91: 625-632.
https://doi.org/10.1016/j.cherd.2013.01.018
Faghihi F., Mirvakili S.M. Shut-down margin study for the next generation VVER-1000 reactor including 13 × 13 hexagonal annular assemblies. Annals of Nuclear Energy 2011; 38: 2533-2540.
https://doi.org/10.1016/j.anucene.2011.07.008
Kalkhoran O.N., Minuchehr A., Shirani A.S., Rahgoshay M. Full scope thermal-neutronic analysis of LOFA in a VVER-1000 reactor core by coupling PARCS v2.7 and COBRA-EN. Progress in Nuclear Energy 2014; 74: 193-200.
https://doi.org/10.1016/j.pnucene.2014.03.006
Rahgoshay M., Tilehnoee M.H. Optimizing a gap conductance model applicable to VVER 1000 thermal-hydraulic model. Annals of Nuclear Energy 2012; 50: 263-267.
https://doi.org/10.1016/j.anucene.2012.08.011
Siefken l.J., Coryell E.E., Harvego E.A., Hohorst J.K. MATPRO - A library of materials properties for light water reactor accident analysis, NUREG/CR-6150, Vol. 4, Rev.2. (2001). p. 2.1 - 2.16 and 4.1 - 4.25..
Terrani K.A., Wang D., Ott L.J., Montgomery R.O. The effect of thermal conductivity on the behavior of LWR cores during Loss-of-Coolant Accidents. Journal of Nuclear Materials 2014; 448: 512-519.
https://doi.org/10.1016/j.jnucmat.2013.09.051
Ott L.J., Robb K.R., Wang D. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident condition. Journal of Nuclear Materials 2014; 448: 520-533.
https://doi.org/10.1016/j.jnucmat.2013.09.052
Shuffler C., Trant J., Malen J., Todreas N. Thermal hydraulic analysis for grid supported pressurized water reactor cores. Nuclear Engineering and Design 2009; 239: 1442-1460.
