小型铅-铋冷却快堆提棒事故核热耦合研究
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摘要
基于热工程序COBRA-YT和物理程序SKRTCH-N,利用幵行虚拟机(PVM)平台开发了核热耦合工具:COBRA-YT将冷却剂密度和燃料温度等热工参数传递给物理程序,用以更新截面;SKETCH-N执行物理计算,幵将功率分布反馈给热工程序;最后,应用该耦合程序分析铅-铋冷却快堆的提棒事故。计算结果显示控制棒提起后,功率迅速升高,在1.42s后达到最大值;5s后包壳温度达到峰值1264℃,超出了设计限值。结果表明:在提棒事故后,均一化布置堆芯的安全会在极短时间内受到严重威胁,故该堆芯应采用分区布置。
The coupled tool based on neutronics code SKETCH-N and thermal-hydraulics code COBRA-YT has been developed via Parallel Virtual Machine(PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is used in the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The reactor power increases rapidly and reaches the peak at 1.42 s after the control rod withdrawal. Meanwhile, the cladding temperature reaches the maximum 1264℃, exceeding its design limit. The results achieved so far indicates that the control rod withdrawal accident poses a threat to the core with the same enrichment, and the optimization work on the core zoning scheme should be done
The coupled tool based on neutronics code SKETCH-N and thermal-hydraulics code COBRA-YT has been developed via Parallel Virtual Machine(PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is used in the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The reactor power increases rapidly and reaches the peak at 1.42 s after the control rod withdrawal. Meanwhile, the cladding temperature reaches the maximum 1264℃, exceeding its design limit. The results achieved so far indicates that the control rod withdrawal accident poses a threat to the core with the same enrichment, and the optimization work on the core zoning scheme should be done
引文
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[2]KOUIDRI W T,LETAIM F,BOUCENNA A,et al.Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled3D neutron kinetics thermal-hydraulic system code technique[J].Progress in Nuclear Energy,2015,85:384-390.
[3]KONDO S,TOBITA Y,MORITA K,et al.Current status and validation of the SIMMER-III LMFR safety analysis code[C],Proc.7th International Conference on Nuclear Engineering(ICONE-7),Kyoto,Japan,1999.
[4]ANGELUCCI M,EBOLI M,FORGIONE N,et al.Transient analyses for the MYRRHA-FASTEF reactor by simmer code[C]//16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics,NURETH 2015.American Nuclear Society,2015,6:4675-4688.
[5]VAZQUEZ M,TSIGE-TAMIRAT H,AMMIRABILE L,et al.Coupled neutronics thermal-hydraulics analysis using Monte Carlo and sub-channel codes[J].Nuclear Engineering&Design,2012,250(3):403-411.
[6]辜峙钘,宋勇,王刚,等.自然循环铅铋冷却快堆失热阱瞬态研究[J].原子能科学技术,2015,49(z1):161-166.
[7]吴宜灿,柏云清,宋勇,等.中国铅基研究反应堆概念设计研究[J].核科学与工程,2014(2):201-208.
[8]GEIST A,BEGUELIN A,Dongarra Jet al.PVM 3 User’s Guide and Reference Manual Technical Report:ORNL/TM-12187[R].Oak Ridge National Laboratory,1994.
[9]ZIMIN V G,ASAKA H,ANODA Y.SKETCH-N:a nodal neutron diffusion code for solving steady-state and kinetics problems.Model description and user’s guide[R].JAERI report Tokyo,Japan,2001.
[10]ASAKA H.Coupling of the thermal-hydraulics TRACcodes with 3D neutron kinetics code SKETCH-N[C].Workshop proceedings of Advanced thermal-hydraulic and neutronic codes:Current and future applications.Barcelona,Spain:NEA/OECD,2000.
[11]MARLEAU G,HéBERT A,ROY R.A user guide for DRAGON 3.06[R].Report IGE-174 Rev,2008.
[12]高新力,田永红,左嘉旭,等.铅铋合金冷却快堆PBWFR子通道参数敏感性研究[J].电网与清洁能源,2016,32(6):131-135.
[13]GUO C,LU D,ZHANG X,et al.Development and application of a safety analysis code for small Lead cooled Fast Reactor SVBR 75/100[J].Annals of Nuclear Energy,2015,81:62-72.