FMSR: A code system for in-core fuel management calculation of aqueous homogeneous solution reactor

Aqueous homogeneous solution reactor is a promising concept for the production of medical isotopes. But some characteristics of aqueous solution reactors, such as no traditional assembly in the core, the gas bubbles’ generation in fuel solution, isotopes distillation, unstructured geometry, strong anisotropic scattering, etc., make the fuel management calculation very complicated. This study establishes a suitable calculation model for aqueous homogeneous solution reactors and developed an in-core fuel management calculation code FMSR (Fuel Management for Solution Reactors) based on the 3D transport solver DNTR. Numerical results indicate that FMSR can be used for the fuel management calculation of homogeneous aqueous solution reactor as a trial.     

界面流法计算反应堆六角形燃料组件中子通量密度分布

利用界面流法计算两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。     

基于任意三角形网格的中子扩散变分节块法

传统的基于矩形和六角形几何的堆芯计算程序已不适用于具有复杂几何的新型反应堆堆芯计算,本文开展了基于任意三角形网格的多群中子扩散变分节块方法研究。首先,采用ANSYS软件对计算区域进行三角形网格剖分,并利用坐标变换将任意三角形变换为正三角形;其次,采用Galerkin变分技术建立包含节块中子平衡方程的泛函,将三角形节块内变量利用正三角形内正交基函数进行展开;最后,利用变分原理,获得中子通量密度与节块边界上分中子流的响应关系,并基于传统的源迭代法对其进行求解。基于上述理论模型开发了程序TriVNM,并采用不同几何基准题进行了验证。结果表明,TriVNM计算的堆芯keff和归一化功率分布与参考解吻合较好,该计算方法适用于复杂几何堆芯扩散计算。     

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。     

溶液堆燃料管理计算方法初步研究与程序研制

溶液型医用同位素生产堆的核燃料呈流动的水溶液形式.堆芯呈非结构、强各向异性散射,运行过程中会产生大量气体.针对堆芯燃料管理计算需要在线提取核素等特点,基于以三角形节块S_N方法为模型的中子输运计算程序DNTR,开发了溶液堆堆芯燃料管理计算程序FMSR,并利用该程序对溶液堆进行了模拟分析.结果表明,FMSR程序可在溶液堆堆芯燃料管理计算中试用.     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

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。     

球床式高温气冷堆初次临界物理计算的蒙特卡罗方法模型分析

对HTR-10初次临界的几何模型进行了对比和分析,运用基于蒙特卡罗方法的MCNP4B和TRIPOLI-4．3程序描述了高温气冷堆的包缀燃料颗粒在燃料球内的随机分布以及燃料球和石墨球在堆芯的随机混合分布应用TRIPOLI-4．3对HTR-10进行了初次临界物理计算,并且与已有的MCNP4B的计算结果进行了比较结果表明：基于蒙特卡罗方法的MCNP4B和TRIPOLI-4．3程序,采用适当的几何描述方式可以用手球床式高温气冷堆的初次临界堆芯物理计算．     

Application of continuous-energy Monte Carlo code as a cross-section generator of BWR core calculations

A continuous-energy Monte Carlo code is newly applied for the assembly calculations of actual BWR core analysis. Few-groups cross-sections and related constants (kinetic parameters) were generated by the continuous-energy Monte Carlo code MVP-BURN, and were tabulated for a core simulator. The commercial BWR, HAMAOKA-3 (1100MWe:BWR-5), was analyzed by a coupled neutronic-thermalhydraulic core simulator based on modified one-group diffusion theory using these assembly constants. The calculated core parameters showed good agreement with the results of the on-line core monitoring system of HAMAOKA-3. Consequently, it was confirmed that the present method is applicable to BWR core production calculations. The present method is a particularly attractive candidate for the analysis of advanced BWR fuel assemblies with exotic geometry and high Gd content, due to the features of the continuous-energy Monte Carlo code, i.e., high accuracy and generalized geometry treatment.     

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.     

Development and validation of a fast reactor core burnup code – FARCOB

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X–Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.     

MCOR – Monte Carlo depletion code for reference LWR calculations

The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged.     

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together.Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary.The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.     

复杂几何燃料组件的参数计算

利用加拿大蒙特利尔大学研制的DRAGON程序对反应堆复杂几何组件进行参数计算，并通过压水堆柱状元件基准问题、MTR型反应堆板状元件基准问题和其他不同几何形状的燃料组件进行校核计算。结果表明：DRAGON程序可用于多种复杂几何燃料组件参数的计算，且具有良好的计算精度。      

Development of discrete ordinates code supporting unstructured tetrahedral mesh and applied in neutronics analysis for the Korea Helium Cooled Ceramic Reflector Test Blanket Module

The discrete ordinates code under development by KAERI uses an unstructured tetrahedral mesh, and thus it can be applied to solve the radiation transport in a complicated geometry. In addition, the geometry modeling process has become much easier because computational tetrahedral meshes are generated based on the CAD file by Gmsh.As our first phase of applying the code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. Visualization of the fluxes and associated reaction rates in the whole system with a single run is one of the merits of a deterministic method and is very useful for checking hot spots.     

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。     

Evaluation of fluid flow in the lower core of a PWR with Code_Saturne

In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. The first spacer grid region is of particular interest, as fuel rod fretting has sometimes been observed at that level. Entry conditions depend on the geometry of the lower core plate and of the assembly nozzles. Distribution of flow in the downcomer and lower plenum is also a factor. A series of calculations are thus run with the incompressible Navier–Stokes solver, Code_Saturne, using a classical RANS turbulence model. The first calculations involve a global geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate and the fuel rod assemblies above it cannot be well represented within a practical mesh size, so that a head loss model is used. Different types of assemblies can be represented through different head loss coefficients. We make full use of Code_Saturne’s non-conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Steady-state or near steady-state results are obtained, which may be used as realistic entry conditions for full core calculations at assembly width resolution, and beyond those, mechanical strain calculations. We are especially interested in more detailed flow conditions and in the lower core area, so as in the future to quantify vibrational input. This requires a much higher resolution, which is limited to a scale of a few assemblies for practical reasons. At this scale, most of the features of the fuel rods, nozzles, and guide tubes are represented, though the geometry of the spacer grids is still much simplified, and details such as debris-trapping grids are ignored. Different meshes are used for different fuel types. For the moment, a constant velocity upstream of the lower core plate is used as an inlet condition. We have also built a small lower fuel rod assembly mock-up (1/5 scale 7 × 7 tube, 3 × 3 assemblies) with which we plan to obtain detailed flow information, and better qualify the use of our CFD codes with regards to this type of application.     

Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark

As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified by Cavarec et. al. [1]. This problem, known as C5G7 MOX Benchmark, is described in the Benchmark Specification [2] and comprises two cases — two and three-dimensional geometry. There are four fuel assemblies — two with MOX fuel, the other two with UO₂ fuel. Each fuel assembly is made up of a 17×17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group transport-corrected isotropic scattering cross-sections for U0₂, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper preset is the methodology employed in solving the C5G7 MOX Fuel Assembly Problem using the transport code HELIOS.     

Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations

Three-dimensional pin-by-pin core analysis is considered to be a candidate for the next-generation BWR core calculation method. In our previous study, the applicability of the transport and burnup calculations for a three-dimensional pin-by-pin BWR core analysis was investigated. However, the thermal-hydraulics calculation has not yet been studied in this framework. In the conventional core analysis code, the bundlewise thermal-hydraulics calculation is adopted. In the actual core analysis, the power distribution inside a fuel assembly is tilted at the region adjacent to a control blade or the core peripheral region. In these regions, the consideration of the subchannel-wise void distribution has an impact on the fission rate distribution. Therefore, an evaluation of the detailed void distribution inside an assembly, i.e., the incorporation of the subchannel wise void distribution, is desirable for the pin-by-pin BWR core analysis. Although several subchannel analysis codes have been developed, these subchannel analysis codes generally require a large computational effort to estimate the subchannel-wise void distribution in a whole BWR core. Therefore, to analyze a whole BWR core within a reasonable computation time, it was necessary to apply a fast subchannel analysis code. In this paper, a quick subchannel analysis code dedicated to pin-by-pin BWR core analysis is newly developed, and the void distribution of the present subchannel analysis code is compared with the prevailing subchannel analysis code NASCA using three-dimensional single-assembly geometries. Since the present subchannel analysis code is used for a coupled neutronics/thermal-hydraulics analysis, the results of the coupling calculation are also compared with those of NASCA. The calculation result indicates that the void distribution difference between NASCA and the present subchannel analysis code is slightly less than 10%. This result indicates that the prediction accuracy of the present subchannel analysis code will be reasonably appropriate for a pin-by-pin BWR core analysis. Furthermore, the results show that the calculation time of the present subchannel analysis code is only 10 min for a hypothetical three-dimensional ABWR quarter-core geometry using a single CPU. This calculation time is sufficient for a pin-by-pin BWR core analysis.     

Reconstruction of cell homogenized macroscopic cross sections for analyzing fast and thermal coupled cores using the SRAC system

A fast and thermal neutron coupled core adopts blanket fuel assemblies with zirconium hydrides in the core for negative coolant void reactivity. Conventional neutronics calculation methods have been developed for analysis of a fast core or thermal core, in which the coarse-group macroscopic cross sections of fuel assemblies are prepared without including the effect of the surrounding fuel assemblies. However, such methods are not adequate for analyzing fast and thermal neutron coupled cores where the intra-assembly and inter-assembly heterogeneity effects must be precisely taken into account. Recently, a concept of reconstruction of cell homogenized macroscopic cross sections has been proposed to take into account effects of inter-assembly heterogeneities on macroscopic cross sections used in the reactor core analysis and successfully applied based on a Monte Carlo method. In the present study, a reconstruction method of cell homogenized coarse-group macroscopic cross section for analyzing fast and thermal coupled cores is developed based on a deterministic neutronics calculation code system, SRAC. Three types of fixed source calculations for unit assembly cell geometry are performed independently of the specific core layouts and their results are combined with the results of core analysis to produce cell homogenized coarse-group macroscopic cross sections. Numerical results show that the heterogeneity effects can be adequately reflected in the reconstructed macroscopic cross sections with the proposed method. When the number of energy groups is small, the proposed method gives poor results in the transitional energy groups from resonance to thermal energy. Therefore, it is necessary to increase the number of energy groups in this energy range.