SARAX-FXS程序针对非均匀快堆堆芯计算中的改进及应用
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摘要
SARAX-FXS程序是基于确定论方法,适用于快谱堆芯组件能谱、均匀化参数计算的程序。由于快堆中组件空间自屏的非均匀效应不可忽视,本文将基于一维圆柱、平板几何的碰撞概率方法加入SARAX-FXS模块,并以等效一维模型计算组件的均匀化参数。为保证能群归并前后的核反应率守恒,在组件计算中引入超级均匀化(SPH)因子修正截面。采用快堆基准题MET-1000对程序的计算结果进行验证,结果表明,与参考解相比,SARAX-FXS的一维计算模块具有较高的精度,特征值计算相对偏差在100~200pcm之间。堆芯计算结果显示,引入SPH因子可提高特征值计算的精度约300pcm,功率分布的均方根误差可从约3%下降至约1%。
The SARAX-FXS code is a calculation program based on the deterministic method.It has the capabilities of calculating neutron spectra and homogenized cross sections of fast reactor analysis.Since the heterogeneous effect within the assembly was not negligible in fast reactor system,the collision probability method(CPM)with onedimension slab and cylinder geometry was introduced into the SARAX-FXS code to obtain the homogenized cross sections with this one-dimension geometry.Meanwhile,in order to conserve the reaction rate after energy group condensation,the super-homogeneous(SPH)factor was also implemented in the calculation.To verify the development of the code,the MET-1000 benchmark loaded with metallic fuel was chosen as the verification case.Compared with the reference solutions,the eigenvalue calculated by SARAX-FXS with CPM shows the high accuracy with 100-200 pcm relative deviation.Meanwhile,the whole core calculation shows that the application of SPH factor could enhance the accuracy about 300 pcm and the root-mean-square error of power distribution could decrease from about 3%to 1%.
The SARAX-FXS code is a calculation program based on the deterministic method.It has the capabilities of calculating neutron spectra and homogenized cross sections of fast reactor analysis.Since the heterogeneous effect within the assembly was not negligible in fast reactor system,the collision probability method(CPM)with onedimension slab and cylinder geometry was introduced into the SARAX-FXS code to obtain the homogenized cross sections with this one-dimension geometry.Meanwhile,in order to conserve the reaction rate after energy group condensation,the super-homogeneous(SPH)factor was also implemented in the calculation.To verify the development of the code,the MET-1000 benchmark loaded with metallic fuel was chosen as the verification case.Compared with the reference solutions,the eigenvalue calculated by SARAX-FXS with CPM shows the high accuracy with 100-200 pcm relative deviation.Meanwhile,the whole core calculation shows that the application of SPH factor could enhance the accuracy about 300 pcm and the root-mean-square error of power distribution could decrease from about 3%to 1%.
引文
[1]TONE T.A numerical study of heterogeneity effects in fast reactor critical assemblies[J].Journal of Nuclear and Technology,1975,12(8):467-481.
[2]KAMEI T,YOSHIDA T.Fuel pin and subassembly heterogeneity effect on neutronics properties of fast power reactor[J].Journal of Nuclear and Technology,1981,18(1):74-76.
[3]CHIBA G.A combined method to evaluate the resonance self shielding effect in power fast reactor fuel assembly calculation[J].Journal of Nuclear and Technology,2003,40(7):537-543.
[4]郑友琦,吴宏春,杜夏楠,等.快堆中子学计算程序NECP-SARAX的改进与验证[J].强激光与粒子束,2017,29(5):116-121.ZHENG Youqi,WU Hongchun,DU Xianan,et al.Improvement and verification of the fast reactor neutronics code NECP-SARAX[J].High Power Laser and Particle Beams,2017,29(5):116-121(in Chinese).
[5]杜夏楠,曹良志,郑友琦.基于点截面的快堆组件均匀化参数计算方法[J].强激光与粒子束,2017,29(1):5-10.DU Xianan,CAO Liangzhi,ZHENG Youqi.Method of generating homogenized fast reactor assembly constants based on point-wise cross sections[J].High Power Laser and Particle Beams,2017,29(1):5-10(in Chinese).
[6]HADER J S,SHEMON E R,LEE C H.Initial verification of heterogeneous cross sections in MC2-3/PROTEUS[C]∥American Nuclear Society Winter Meeting.Anaheim,CA:[s.n.],2014.
[7]JARRETT M G,SHEMON E R,LEE C H.Heterogeneous multigroup cross sections for fast reactor calculations with MC2-3/PROTEUS[C]∥PHYSOR 2016.Sun Valley:[s.n.],2016.
[8]HEBERT A.A reformulation of the transporttransport SPH equivalence technique[C]∥7th ICMSNSE.Canada:[s.n.],2015.
[9]NIKITIN E,FRIDMAN E,MIKITYUK K.On the use of the SPH method in nodal diffusion analyses of SFR cores[J].Annals of Nuclear Energy,2015,85:544-551.
[10]谢仲生,邓力.中子输运理论数值计算方法[M].西安:西北工业大学出版社,2005.
[11]OECD/NEA.Benchmark for neutronic analysis of sodium-cooled fast reactor cores with various fuel types and core sizes[R].[S.l.]:Nuclear Energy Agency and Organization for Economic Co-operation and Development,2016.
[12]Computational Reactor Physics Group.OpenMC documentation,Release 0.7.1[EB/CL].https:∥github.com/mit-crpg/openmc.
[2]KAMEI T,YOSHIDA T.Fuel pin and subassembly heterogeneity effect on neutronics properties of fast power reactor[J].Journal of Nuclear and Technology,1981,18(1):74-76.
[3]CHIBA G.A combined method to evaluate the resonance self shielding effect in power fast reactor fuel assembly calculation[J].Journal of Nuclear and Technology,2003,40(7):537-543.
[4]郑友琦,吴宏春,杜夏楠,等.快堆中子学计算程序NECP-SARAX的改进与验证[J].强激光与粒子束,2017,29(5):116-121.ZHENG Youqi,WU Hongchun,DU Xianan,et al.Improvement and verification of the fast reactor neutronics code NECP-SARAX[J].High Power Laser and Particle Beams,2017,29(5):116-121(in Chinese).
[5]杜夏楠,曹良志,郑友琦.基于点截面的快堆组件均匀化参数计算方法[J].强激光与粒子束,2017,29(1):5-10.DU Xianan,CAO Liangzhi,ZHENG Youqi.Method of generating homogenized fast reactor assembly constants based on point-wise cross sections[J].High Power Laser and Particle Beams,2017,29(1):5-10(in Chinese).
[6]HADER J S,SHEMON E R,LEE C H.Initial verification of heterogeneous cross sections in MC2-3/PROTEUS[C]∥American Nuclear Society Winter Meeting.Anaheim,CA:[s.n.],2014.
[7]JARRETT M G,SHEMON E R,LEE C H.Heterogeneous multigroup cross sections for fast reactor calculations with MC2-3/PROTEUS[C]∥PHYSOR 2016.Sun Valley:[s.n.],2016.
[8]HEBERT A.A reformulation of the transporttransport SPH equivalence technique[C]∥7th ICMSNSE.Canada:[s.n.],2015.
[9]NIKITIN E,FRIDMAN E,MIKITYUK K.On the use of the SPH method in nodal diffusion analyses of SFR cores[J].Annals of Nuclear Energy,2015,85:544-551.
[10]谢仲生,邓力.中子输运理论数值计算方法[M].西安:西北工业大学出版社,2005.
[11]OECD/NEA.Benchmark for neutronic analysis of sodium-cooled fast reactor cores with various fuel types and core sizes[R].[S.l.]:Nuclear Energy Agency and Organization for Economic Co-operation and Development,2016.
[12]Computational Reactor Physics Group.OpenMC documentation,Release 0.7.1[EB/CL].https:∥github.com/mit-crpg/openmc.