六角形轻水堆组件中子通量密度分布的计算

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。     

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

研究利用穿透概率法求解二维六角形轻水堆燃料组件内中子通量密度分布。子区内中子源采用线性分布,子区表面通量密度在方向上采用简化6P1近似。提出了六角形组件周边水隙的处理方法。根据提出的模型,编制了TPHEX－C程序,并对六角形组件进行了计算,结果与蒙特卡罗方法计算的结果符合良好。     

轻水反应堆燃料组件中子通量和功率分布的计算程序

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。     

界面流方法二维六角形组件计算

讨论了用界面流方法计算二维六角形组件中子通量分布。从积分输运方程出发,导出了一种简便的数学模型,在子区内采用平源通量近似,并假设中子发射和散射为各向同性。在子区表面上,中子通量的空间分布为常数,中子角通量分布通过伴随勒让特多项式展开表示,采用DP_1近似。推导出界面流方程组,给出了泄漏、穿透几率矩阵及其矩阵元素的表达式及计算方法。根据提出的数学模型,编制了TPHEX程序,对二维六角形组件进行了计算,本程序可用于水堆六角形燃料组件计算。     

二维轻水堆燃料组件中子通量分布的计算

本文讨论了用穿透几率法计算二维轻水堆燃料组件中子通量的分布,提出了一种简易的计算模型。在子区内中子通量采用线性分布,子区表面上采用P_1近似角分布和线性空间分布,对展开系数导出了简便的表达式,即用表面上的出射和入射中子流来决定,并在迭代过程中逐步精确化。因而减少了求解的未知量,简化了计算。根据提出的模型,编制了二维TPM2D计算程序。对轻水堆的一些组件基准问题作了计算。计算结果与S_N、节块S_N以及积分输运理论等方法进行了比较,其结果符合得很好。本程序可用于轻水堆燃料组件的计算。     

快堆燃料组件内精细功率分布的计算

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。     

穿透几率法求解X-Y几何多群中子积分输运方程

本文介绍应用穿透几率法求解X-Y几何多群中子输运方程的程序—TPXY。把系统分成若干子区,子区内部中子通量呈线性分布,子区表面通量角分布采用4P_1近似,空间呈线性分布。对一系列组件基准问题作了校验计算,计算结果与S_N,碰撞几率及其它穿透几率方法的结果进行了比较,均符合良好。本程序可用于轻水堆组件计算。     

二维六角形轻水堆燃料组件中子通量分布的计算

介绍利用穿透概率法求解二维六解形几何多群中子积分输运方程。子区内中子源及通量采用线性分布,子区表面通量在方向上采用简化６Ｐ１近似。根据提出的模型,编制了ＴＰＨＥＸ－Ｂ程序,并对一些轻水堆六解形组件问题做了计算,计算结果与ＭＣ结果进行了比较,符合良好。本程序可用于六解形轻水堆燃料组件计算。     

三维六角形组件压水堆堆芯燃料管理计算及程序系统研究

介绍所研制的WWER型压水堆堆芯燃料管理计算程序系统TPFAP-H／CSIM-H，六角形组件均匀化计算程序TPFAP-H是在压水堆正方形组件程序TPFAP的基础上，采用穿透概率法与响应矩阵方法相结合计算六角形组件内中子能谱分布，并考虑六角形栅元特点改造开发而成的CSIM-H是以先进六角形节块扩散程序为基础．参照SIMULATE程序功能而研制的物理-热工水力耦合的三维六角形节块PWR堆芯燃料管理程序两者通过接口程序LINK连接起来，可以考虑燃耗，功率、慢化剂密度变化．控制棒、氙等参数的多种反馈效应对IAEA的WWER-1000型Kalinin核电厂基准问题的校算的结果表明，临界硼浓度、功率和燃耗分布等结果与国际各研究机构的结果吻合良好，偏差均在工程要求之内。     

压水堆六角形燃料组件均匀化 计算软件包TPFAP-HEX

介绍了所研制的具有工程实用价值的压水堆六角形燃料组件均匀化计算软件包。该组件中子空间能谱的计算采用穿透概率法与响应矩阵法相结合的方法，在六角形几何内求解中子积分输运方程。在此方法中，栅元内中子源采用空间线性或二次近似，栅元表面中子通量密度角分布采用简化6P     

中子输运计算界面流方法的微扰计算

以六角形几何中子积分输运计算界面流算法及其相对应的数学共扼方程计算为基础,利用微扰原理计算了当反应堆六角形组件中栅元核参数发生微量变化时系统反应性的变化。计算结果表明,本文所开发的基于六角形几何中子积分输运算法的微扰计算方法是正确的。     

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

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.     

Transmission probability method based on triangle meshes for solving unstructured geometry neutron transport problem

In the advanced reactor, the fuel assembly or core with unstructured geometry is frequently used and for calculating its fuel assembly, the transmission probability method (TPM) has been used widely. However, the rectangle or hexagon meshes are mainly used in the TPM codes for the normal core structure. The triangle meshes are most useful for expressing the complicated unstructured geometry. Even though finite element method and Monte Carlo method is very good at solving unstructured geometry problem, they are very time consuming. So we developed the TPM code based on the triangle meshes. The TPM code based on the triangle meshes was applied to the hybrid fuel geometry, and compared with the results of the MCNP code and other codes. The results of comparison were consistent with each other. The TPM with triangle meshes would thus be expected to be able to apply to the two-dimensional arbitrary fuel assembly.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

A co-ordinate system for reactor physics calculations in hexagonal geometry

A method for generating all the geometric information concerning typical reactor physics calculations for a basically hexagonal reactor core or its sector involving any of the possible symmetries is presented. The geometrically allowed symmetries for regular hexagons are discussed. The approach is based on the choice of a suitable co-ordinate system, viz. one using three coplanar (including one redundant) axes, each at 120° with its cyclically preceding one. A code named KEKULE' is developed for a 2-D, finite difference, one-group diffusion analysis of a hexagonal core using the approach. It can cater to a full hexagonal core as well as to any symmetric sectorial part of it. The main feature of the code is that the input concerning geometry is a bare minimum. It is hoped that the approach presented will be useful even for the calculations for hexagonal fuel assemblies.     

Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).