CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

Wahid Luthfi, Surian Pinem

DOI: http://dx.doi.org/10.17146/tdm.2020.22.3.5955

Abstract


The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.

Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006


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References


  1. Zheng Y., Wu H., Cao L., Jia S. Economic evaluation on the MOX fuel in the closed fuel cycle. Sci. Technol. Nucl. Install. 2012. 2012
  2. Reda S.M., Gomaa I.M., Bashter I.I., Amin E.A. Effec of MOX Fuel and the ENDF / B-VIII on the AP1000 Neutronics Parametrs Calculation by Using MCNP6. 2019. 34(4):325–35.
  3. Selim H.K., Amin E.H., Roushdy H.E. Rod ejection accident analysis for AP1000 with MOX/UOX mixed core loading. Ann. Nucl. Energy. 2017. 109:385–95.
  4. Mouginot B., Leniau B., Thiolliere N., Bidaud A., Courtin F., Doligez X., et al. MOX fuel enrichment prediction in PWR using polynomial models. Ann. Nucl. Energy. 2014. 85:812–9.
  5. Fetterman R.J. Annals of Nuclear Energy AP1000 core design with 50 % MOX loading. Ann. Nucl. Energy. 2009. 36(3):324–30.
  6. International Atomic Energy Agency Annual report 2003. IAEA Nucl. Energy Ser. 2003.(December):1–24.
  7. Salam M., Hah C.J. Comparative study on nuclear characteristics of APR1400 between 100% MOX core and UO2 core. Ann. Nucl. Energy. 2018. 119:374–81.
  8. Elsawi M.A., Hraiz A.S.B. Benchmarking of the WIMS9/PARCS/TRACE code system for neutronic calculations of the Westinghouse AP1000TM reactor. Nucl. Eng. Des. 2015.
  9. El Ouahdani S., Boukhal H., Erradi L., Chakir E., El Bardouni T., Hajjaji O., et al. Monte Carlo analysis of KRITZ-2 critical benchmarks on the reactivity temperature coefficient using ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries. Ann. Nucl. Energy. 2016. 87:107–18.
  10. Sembiring T.M., Pinem S., Liem P.H. Validation of full core geometry model of the NODAL3 Code in the PWR transient benchmark problems. Tri Dasa Mega. 2015. 2015:141–8.
  11. Okumura K., Kugo T., Kaneko K., Tsuchihashi J. SRAC2006: A Comprehensive Neutronics Calculation Code System. Tokai:Japan Atomic Energy Agency; 2007.
  12. OECD Nuclear Energy Agency Pressurised Water Reactor MOX / UO2 Core Transient Benchmark Final Report NEA Nuclear Science Committee Working Party on Scientific Issues of Reactor Systems Pressurised Water Reactor MOX / UO 2 Core Transient Benchmark. 2014.
  13. Pinem S., Sembiring T.M., Liem P.H. The verification of coupled neutronics thermal-hydraulics code NODAL3 in the PWR rod ejection benchmark. Sci. Technol. Nucl. Install. 2014. 2014:1–9.
  14. Liem P.H., Pinem S., Sembiring T.M., Tran H. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3). Nucl. Sci. Technol. 2016. 6(1):1–13.
  15. Pinem S., Sembiring T.M., Liem P.H. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR). Sci. Technol. Nucl. Install. 2016. 2016:1–11.
  16. Purdue University OECD/NEA AND U.S. NRC PWR MOX/UO2 CORE TRANSIENT BENCHMARK [Accessed: 7 July 2020]. Available from: https://engineering.purdue.edu/PARCS/MOX_Benchmark.


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