共查询到18条相似文献,搜索用时 234 毫秒
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六角形轻水堆组件中子通量密度分布的计算 总被引:2,自引:0,他引:2
介绍利用穿透概率法求解二维六角形轻水堆燃料组件中子通量密度分布。子区内中子源及通量密度在空间上采用二次分布 ,子区表面通量密度在空间上采用平通量密度分布 ,在方向上采用简化 6P1近似。根据提出的模型 ,编制了TPHEX D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MC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。 相似文献
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三维六角形节块多群中子扩散程序NDHEX 总被引:2,自引:2,他引:0
本文介绍用DIF3D (NOD)求解二、三维六角形几何系统下中子扩散方程的理论模型及数值计算方法。六角形节块内的中子通量密度分布采用高次多项式近似表示,最后导出通量矩方程及偏流的响应矩阵方程。应用粗网再平衡和渐近源外推方法加速收敛。参考此方法编制了计算程序NDHEX,并对一些六角形基准问题进行了计算。结果表明:NDHEX的计算结果与DIF3D(NOD)的计算结果符合很好;与差分程序相比,具有更高的精度与计算效率。它可用于快堆计算。 相似文献
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本文介绍用穿透几率法计算二维轻水堆燃料组件内中子通量分布的两种计算模型和程序.在子区内及表面上中子通量采用线性空间分布近似,子区表面上角通量分别采用准 DP_1和 QP_1近似。对一些轻水堆组件基准问题作了验证计算。计算结果与 S_N、节块 S_N 以及积分输运理论等方法进行比较,其结果符合良好。这些程序可用于轻水堆燃料组件的计算。 相似文献
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提出了一种在三维六角形几何节块内数值求解中子扩散方程的节块法该方法把节块内各群中子注量分布用解析基函数近似展开为了改善节块耦合关系.提出了,一种新的节块边界条件:面平均偏流零次矩和一次矩同时保持连续。此外.将响应矩阵技术应用于迭代求解过程,使得该方法具有较高的计算效率基于本文提出的模型,发展了三维六角形组件中子扩散计算程序FEMHEX。通过对二维、三维VVER基准问题校验计算表明,该方法能高效.准确的给出有效增殖系数以及节块功率分布。 相似文献
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提出了一种在二维三角形几何内数值求解中子扩散方程的节块方法.节块内的各群通量分布用解析基函数近似展开,节块之间采用面偏流零次矩和一次矩进行耦合;给出了三角形几何下的节块扫描方案;采用响应矩阵技术进行迭代求解,开发了二维三角形组件中子扩散计算程序ABFEM-T.通过基准问题的校验计算,表明该方法能准确地给出有效增值系数及节块功率分布,可用于复杂的非结构几何区域的中子扩散问题的求解. 相似文献
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堆芯热通道因子是堆芯热工设计及安全分析的一项重要参数,确定热通道因子需用中子学计算给出较准确的燃料组件内元件棒功率分布。在三维六角形几何节块扩散理论基础上,使用多项式重构的方法计算节块内中子通量密度分布和功率密度分布。针对快堆六角形燃料组件的特点,用小六角形积分的方法计算组件内元件棒功率,得到组件内各元件棒功率分布。在NAS程序基础上,编制了元件棒功率分布计算模块NAS PIN。通过与蒙特卡罗程序的校验可发现,二者计算结果符合较好,计算精度可满足工程设计的需要。 相似文献
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基于中子积分输运理论,应用综合界面流和碰撞几率技巧的块方法,导出了处理三区非均匀栅元结构的二维(X-Y)几何多群中子输运问题的数值模型。即对于由若干栅元组成的按X-Y几何排列的堆芯结构,对每一类栅元剖分为圆柱形元件(如燃料棒、控制棒、可燃毒物棒等)、包壳和慢化剂三个均匀区,用碰撞几率(CP或PIJ)方法计算各区的中子通量分布;对于相邻栅元用DP1近似的中子流来耦合;因此,块方法具有精度高、速度快、能灵活处理各种几何问题的优点,是目前动力堆组件计算最有前途的方法之一。基于块方法基本理论,发展了三区栅元模型,导出了计算方法,编制了FORTRAN计算机程序。为验证其精度和适用性,对两个例题进行了计算,并与其它程序的计算结果进行了比较,证明功率分布和本征值均符合较好。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):983-991
In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail. 相似文献
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P. Mohanakrishnan 《Progress in Nuclear Energy》2004,45(2-4):125-131
Interface current approach to solution of neutron transport equation has been earlier used for LWR lattice problems. The analysis of the C5MOX benchmark is a opportunity to test its applicability to heterogeneous reactor problem. Computer code COHINT, which incorporates a routine for solution of neutron transport equation in X-Y geometry by interface current approach was used for this analysis by representing the fuel rod as a square one. Region interface angular fluxes are represented by a four-term expansion. It is found that the average pin power error is about 2.28 % (peak pin error 4.1 %) relative to reference calculations. Further improvement is possible by introduction of the capability to represent circular rods with in a square cell in COHINT. 相似文献
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六边形燃料组件在液态金属冷却快堆尤其是钠冷快堆中被广泛应用,针对这类堆型的设计与安全分析需要对堆芯中子通量与中子流进行三维全堆芯耦合计算。经过多年发展,目前已有多种解析节块法、积分节块法、节块展开法等先进节块法能在笛卡尔坐标系下较为精确求解多维中子扩散方程。本文通过径向半解析节块法耦合轴向高阶节块展开法的综合节块方法开发了反应堆三维中子物理计算软件SA HNHEX,并对VVER 440二维、三维基准题进行建模与仿真计算。计算结果与参考值符合较好,初步验证了使用该方法进行反应堆堆芯中子扩散计算的正确性。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(7):348-357
Abstract A method has been developed that effectively estimates the detailed distribution of power generation in the fuel or blanket assemblies in nuclear reactors. A two-dimensional, one-group diffusion model is applied to a region of homogeneous composition enclosed in a contour devoid of concavity viewed from outside. The diffusion equation is reduced to the form of Helmholtz equation, and a non-homogeneous boundary condition of Dirichlet or Neumann type is given on the contour, using neutron fluxes previously obtained in coarse mesh diffusion criticality calculations covering the whole reactor. This boundary value problem in two-dimensional space is made to yield a solution in the form of a potential due to a single or double layer. The method is applied to a hexagonal cell of a fast reactor. The results of calculation are amply accurate in comparison with the corresponding values from the usual fine-mesh diffusion scheme and with much shorter computing time. 相似文献
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The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor. 相似文献