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界面流法计算反应堆六角形燃料组件中子通量密度分布 被引量:4

Calculation of Flux Distribution in Hexagonal LWR Fuel Assembly by Interface Current Method
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摘要 利用界面流法计算两维六角形轻水堆燃料组件内中子通量密度分布。子区内中子源在空间上采用二次分布近似,还考虑了六角形组件周边水隙对组件内中子通量密度的影响。根据提出的模型,编制了TPHEX-E程序,并对一些轻水堆六角形组件问题作了计算,计算结果与蒙特卡罗方法计算结果进行了比较,符合良好。本程序可用于六角形轻水堆燃料组件计算。 This paper describes the model of calculating the flux distribution in two dimensional hexagonal geometry light water reactor fuel assembly by interface current method.The source within the meshes is assumed to be quadratic dependent on x and y coordinates.And the wa ter gap sur-rounding assembly is also considered.Based on proposal model,the code TPHEX-E is encoded and some hexagonal light water reactor assembly problems have been tested.The calculated results are compared with those of Monto Carlo method.They are all in good agreement.The code TPHEX-E can be used to the hexagonal geometry assembly calculation of LWR.
出处 《核动力工程》 EI CAS CSCD 北大核心 2002年第2期87-92,共6页 Nuclear Power Engineering
关键词 界面流法 计算 反应堆 六角形燃料组件 界面流 中子通量密度 分布 轻水堆 Hexagonal geometry Fuel assembly Interface current Neutron flux
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  • 1王梓坤,常用数学公式大全,1991年
  • 2咸春宇,硕士学位论文,1990年
  • 3黄群英,硕士学位论文,1989年
  • 4简靖文,硕士学位论文,1987年
  • 5赵春雷,硕士学位论文,1984年
  • 6陈仁济,中子碰撞概率方法及其应用,1981年
  • 7张颖,西安交通大学学报
  • 8谢仲生,核反应堆物理数值计算,1996年
  • 9江新标,陈达,张颖.超热中子束慢化材料的特性分析[J].核动力工程,2000,21(5):435-438. 被引量:5

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  • 1张颖,陈伟,陈立新.中子输运计算界面流方法的数学共扼方程[J].核动力工程,2005,26(2):97-101. 被引量:1
  • 2Casal J J, Stamm'ler R J J, Villarino E A, et al. HELLOS: Geometric Capabilities of a New Fuel-Assembly Prog- ram[C]. Proc International Topical Meeting on Advances in Mathematics, Computations, and Reactor Physics, Pittsburgh, April 28-May 2, 1991.
  • 3Leszczynski F, Lopez Aldama D, Trkov A. WIMS-D Library Update Final Report of a Coordinated Research Project [R]. IAEA, Vienna, 2006.
  • 4Marleau G, Hebert A, Roy R. A User Guide for Dragon Version4[R]. IGE-294, Technical Report, Ecole Poly- technique de Montreal, 2008.
  • 5Sidorenko V D, Aleshin S S. Verification of 3-D Gene- ration Code Package for Neutronic Calculations of VVERs[C]. 10th AER Symposium on VVER Reactor Physics and Reactor Safety, Moscow, Russia, September 18-22, 2000.
  • 6Aleshin S S, Bolshagin S N, Lazarenko A P, et al. Veri- fication ofRRC KI Code Package for Neutronic Calcula- tions of VVER Core with Gd[C]. 11th AER Symposium on VVER Reactor Physics and Reactor Safety, Csopak, Hungary, September 23-30, 2001.
  • 7Sidorenko V D. Spectral Code TVS-M for Calculation of Characteristics of Cells, Supercells and Fuel Assemblies of VVER-Type Reactors[C]. 5th Symposium of the AER, Dobogoko, Hungary, October 15, 1995.
  • 8Gomin E A, Maiorov L V. The MCU Monte Carlo Code for 3D Depletion Calculation [C]. Proc of Intern Conf on Math. and Comp. Reac. Phys. and Envir. Anal. in Nucl Appl, M&C' 99-Madrid. Madrid, Spain, Sept. 27-30, 1999.
  • 9小斯特西.核反应堆物理学中的变分法[M].杜祥琬,译.北京:原子能出版社,1982.
  • 10Yang W S, Taiwo T A, Khalil H. Solution of the Mathematical Adjoint Equations for an Interface Current Nodal Formulation[J]. Nucl Sci & Eng, 1994,116: 42~54.

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