全谱系中子学计算程序FSAR的研发进展及其共振方法初步验证

Development of Full Neutron Spectrum Code for Advanced Reactor Simulation FSAR and Its Primary Verification of Resonance Method

  • 摘要: 为满足覆盖广域中子能量范围的先进反应堆研发需求,中国核动力研究设计院开发了先进反应堆全谱系中子学计算程序FSAR。FSAR程序是基于确定论两步法计算策略、由二维截面生成计算程序和三维堆芯计算程序组成的中子学计算程序。二维截面生成程序采用超细群条件下的子群方法进行共振自屏效应处理,在共振计算中嵌入特征线方法求解慢化方程以获得精确的共振自屏截面。在截面生成计算中采用特征线方法求解二维组件的中子通量密度,从而获得少群均匀化截面。全谱系先进反应堆中子能量跨度大,不同的中子平均自由程导致了全堆芯局部非均匀性和全局空间耦合效应强烈。为更好地处理上述效应,截面生成计算中采用了超均匀化方法和泄漏修正模型进行均匀化截面修正。堆芯计算程序采用离散纵标法进行三维中子输运计算,采用微观燃耗计算方法模拟堆内全寿期内各核素的消耗和累积。本文采用快谱、中间能谱及热谱条件下的计算问题对FSAR程序共振计算方法进行了初步验证。计算结果表明,对于不同能谱条件下的测试问题,FSAR程序均能够获得与参考解符合较好的有效共振截面,具备良好的全中子能谱共振效应处理能力。

     

    Abstract: The research and development of advanced reactors has been pushed globally in recent decade and promotes the development of reactor physics codes. The neutron spectrum is complicate for some new reactors, especially for the special-use ones. In order to meet the requirements of the advanced reactors, it is necessary to develop the reactor analysis code with full neutron spectrum adapt ability. The full neutron spectrum code for advanced reactor simulation named FSAR has recently been developed at Nuclear Power Institute of China in order to meet the requirements of advanced reactor with large neutron energy range. The deterministic two-step calculation strategy based on the homogenization theory is utilized in FSAR to perform the reactor core neutronics analysis. Firstly, the two-dimensional lattice calculation is performed. The two-dimensional method of characteristics solver in hexagonal geometry for neutron transport calculation and the subgroup method with ultrafine groups for self-shielding calculation are utilized. For the typical assemblies, ultrafine-group cross sections and neutron flux will be determined and the few-group homogenized micro cross sections will be collapsed based on the flux-volume weight method and the principle of conservation of reaction rate. The single assembly problems with reflective boundary condition are used to determine the homogenized cross sections of materials. For better consideration of the strong space coupling in different geometric sizes due to different mean free path of neutrons, the super homogenization method and the leakage model are used as the homogenization technique. For the structural materials and other non-fuel materials, the super cell or assembly models are used to determine the SPH factors and the cross sections. Secondly, based on the multi-group cross sections generated by the lattice calculation, the three-dimensional whole core calculation is carried out by the core simulator. The SN method with triangular grid is applied as the solver of the transport equations. The three-dimensional multi-group neutron transport equation within the triangular prism grid will be calculated with isotropic fission sources and anisotropic scattering sources. The micro-depletion scheme is applied to simulate the core burn up and the Chebyshev rational approximation method is used to solve the depletion equation. In FSAR, the depletion chain containing 21 heavy isotopes and 49 fission products is provided for the advanced reactor core system. Preliminary verification of the resonance method was carried out. Several pin cell problems with different spectrum characteristics were selected as the verification problems. The results for the eigenvalue keff and for the absorption cross sections of 238U and 239Pu were compared to the reference results which were provided by the Monte Carlo code OpenMC. The results indicated that FSAR has good performances in dealing with the resonance self-shielding effect of the full-range spectrum problems.

     

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