Multi-group pin power reconstruction method based on colorset form functions

A multi-group pin power reconstruction method that fully exploits nodal information obtained from global coarse mesh solution has been developed.It expands the intra-nodal flux distributions into nonseparable semi-analytic basis functions,and a colorset based form function generating method is proposed,which can accurately model the spectral interaction occurring at assembly interface.To demonstrate its accuracy and applicability to realistic problems,the new method is tested against two benchmark problems,including a mixed-oxide fuel problem.The results show that the new method is comparable in accuracy to fine-mesh methods.     

A simplified methodology for nuclear waste repository thermal analysis

A simplified model for repository thermal analysis is presented in this paper. The proposed model is to provide a general capability to efficiently calculate the time dependent temperature field in a geologic repository. The model analyzes both horizontal and vertical emplacement of nuclear waste packages. Verification of the code was performed based on the comparison with detailed numerical method-based standard models. The new model’s utility was demonstrated through a case study where a large number of repository-scale thermal analysis calculations is needed.     

Yucca Mountain engineered barrier system corrosion model (EBSCOM)

A revised engineered barrier system model has been developed by the Electric Power Research Institute to predict the time dependence of the failure of the drip shields and waste packages in the proposed Yucca Mountain repository. The revised model is based on new information on various corrosion processes developed by the US Department of Energy and others and for a 20-mm-thick waste package design with a double closure lid system. As with earlier versions of the corrosion model, the new EBSCOM code produces a best-estimate of the failure times of the various barriers. The model predicts that only 15% of waste packages will fail within a period of 1 million years. The times for the first corrosion failures are 40,000 years, 336,000 years, and 375,000 years for the drip shield, waste package, and combination of drip shield and the associated waste package, respectively.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I)

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include:

1. Froth level (or dry-out level) calculation

2. Transition and mixing between convection flow regimes in convective heat transfer

3. Radiant heat transfer between solid walls and flowing gas

4. Heat generation by zirconium-water reaction

5. Crucibilization effect of zirconium-oxide layer

6. Steam starvation effect on zirconium-water reaction.

This code does not calculate motion of fuel rod material but predicts the beginning of relocation. The major affecting models, froth level calculation model, heat transfer model and crucibilization model, are verified through analyses of experiments. This code can be used for thermal hydraulic analysis of a severe accident and fuel damage experiment until significant material relocation occurs.     

Vapor Shield Models for Fusion Reactors Plasma-Facing Components

Radiant heat flux is a dominant mechanism by which energy transfers from the high-temperature core plasma to the interior critical components of the fusion reactor, which result in surface ablation and sever damage to the components. A vapor layer develops at the surface and provides a self-shielding mechanism at the plasma-material interface. Two models for the energy transmission factor through the boundary layer were developed and incorporated in the electrothermal plasma capillary code to predict the effectiveness of these models in surface self-protection. The electrothermal plasma capillary discharge simulates the typical conditions of fusion reactors disruption and quench phase and has been shown to be an adequate technique to evaluate the erosion of plasma-facing component. First model treats the radiant heat transport as it is affected by the variation of the plasma opacity, in which the vapor shield efficiency depends on the plasma optical thickness and the mean plasma opacity. The second model defines the vapor shield by the ratio of the energy reaching the surface to the total radiant energy emitted by the plasma with the inclusion of the plasma kinetic energy. The code can predict the axial and temporal variation of the transmission factor at each time step and mesh point, and predicts the plasma parameters with the effectiveness of the vapor shield at the boundary layer. The code prediction with implementation of both models has been used to compare the results with earlier ones and with some experimental data. Code results are in good correlation with experimentally measured ablation data.     

Modeling reactivity devices for advanced CANDU reactors using the code DRAGON

Full core analysis of typical power reactors generally performed uses few group diffusion theory, it is necessary to generate beforehand, using a lattice code, the required few group cross-sections and diffusion coefficients associated with each region in the core.

For the ACR™ (Advanced CANDU Reactor), the problem is more complex because these reactors contain vertical reactivity devices that are located between two horizontal fuel bundles. The usual calculation scheme relies in this case on a 2D fuel cell calculation to generate the few group fuel properties and on a 3D supercell calculation for the analysis of the reactivity devices present in the core. Because of its complexity, the supercell calculations have usually been performed using simplified fuel geometries. The development of new geometry features in DRAGON and the availability of faster computers have made it possible to improve the 2D cell and 3D supercell models by using explicitly 3D assemblies of clusters to simulate the reactivity devices in CANDU reactors, including the ACR. These studies will thus improve the fine reactor core results by generating more accurate and appropriate reactor databases.

In this paper, we will review the lattice-cell/supercell calculation procedure using the code DRAGON by introducing a new supercell model. The use of such an explicit 3D geometry implies a very fine spatial mesh discretization that can generate a large number of regions leading to problems that cannot be solved by the collision probability (CP) method. The method of characteristics (MoC) is then the only alternative for such cases. A comparison of results using these two methods will also be presented for 3D models with a coarse mesh discretization.     

MCDB Monte Carlo Code with Fast Track Technique and Mesh Tally Matrix for BNCT

The MCDB Monte Carlo Dosimetry Code for Brain Tumors has been developed for Boron Neutron Capture Therapy (BNCT). This code system includes a medical preprocessor, a Monte Carlo dose calculation and a postprocessor. The center point method is used to determine the material and density in each voxel. For dose calculations, a fast particle track technique is developed for the voxel model. This technique is valid for the single-voxel model as well as the multivoxel model. The material matrix and mesh tally matrix are designed for reducing the initialization and tally times. The same dose results as MCNP are achieved. MCDB is faster by a factor of 2:7~3:5 in computational speed with respect to MCNP. Furthermore, MCDB can do parallel computation. It meets the clinical requirements for precision and time.     

A moving subgrid model for simulation of reflood heat transfer

In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance.One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable.A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code.     

Transient failure of zircaloy cladding

A state of the art study of transient failure of zircaloy reveals that a wide range of failure models exist, and their use in fuel analysis depends upon the type of application. Licensing analyses, for example, employ highly simplified criteria that depict cladding rupture in terms of a single response variable, usually the nominal circumferential stress, as function of temperature; while fuel behavior analyses use more detailed models with varying levels of sophistication. Some of these models, however, are ill-suited for detailed fuel rod analysis computer codes because of their dependence on case-specific parameters; others are limited in their utility because of inherent deficiencies in their theoretical treatment of the phenomena involved. Based on this study, a new zircaloy transient failure model is developed that takes account of the important variables involved in transient phenomena, namely, heating rate, strain rate, temperature, true stress, true strain, and prior condition of the cladding material. The model is particularly suited for detailed fuel rod analysis and thus was implemented in the FREY code.     

西安脉冲反应堆在非脉冲瞬态工况下的动态特性研究

西安脉冲堆仪表与控制系统正在进行数字化改造,需要一套动态特性分析程序提供非脉冲瞬态工况下实时变化的功率和燃料温度等参数。本文在以往铀氢锆脉冲堆经典分析程序的基础上,对堆芯物理模型进行了优化和补充,建立了西安脉冲堆动态特性分析模型,开发了可用于非脉冲瞬态工况分析的西安脉冲反应堆动态特性分析程序(XPRDCA),开展了堆上实验,将程序的计算结果与堆上实验结果进行了对比,并研究了燃料温度反应性系数和气隙传热系数对动态特性的影响。结果表明,程序计算结果和堆上实验结果符合较好,采用优化后模型的动态特性分析程序计算速度显著提升。该程序可以用于数字化仪表与控制系统的设计和调试。     

Thermal Fluid-Structure Interaction Analysis for the Upper Structure of LMFRs

This paper describes a thermal fluid-structure interaction analysis code FLUSH that calculates both thermal-hydraulics and thermal structure response at the same time. This code has been developed to evaluate the thermal responses of the upper structures of LMFRs, using two different analysis codes of α-FLOW and FINAS. The heat flux on the boundary surface of the fluid region and the temperature on the boundary of the structure region are exchanged in every iterative cycle as the new boundary conditions, and finally the unified thermal fields are calculated. The different mesh method and the detail thermal radiation model were also developed to apply for the large scale models.

The 2-D model of the basic experiment for the cover gas thermal-hydraulics was calculated to verify this iterative method. The calculated average temperature on the boundary agreed well with the experimental results. The 3-D large scale models of the out-of-pile experiment for MONJU shield plug were also calculated to verify this method. The calculated temperature both in the annulus and the shield plug agreed well with the experiments. These studies showed that this iterative method of FLUSH was very effective for the predictions in the strong coupled thermal fields.     

数值反应堆多物理多尺度耦合技术进展研究

介绍了数值反应堆的基本概念,详细调研了国际上针对数值反应堆开展的研发项目,如轻水堆先进仿真联盟（CASL）、欧洲核反应堆仿真通用平台（NURESIM）和核能先进仿真与建模（NEAMS）项目,总结了多物理耦合及多尺度耦合技术的国内外研究现状,并结合研究现状指出材料腐蚀行为与流动传热、中子物理共同作用下的多物理耦合机理、基于统一网格求解的高保真耦合程序开发是数值反应堆技术发展的重点方向。      

ASTEC and ICARE/CATHARE modelling improvement for VVERs

ASTEC and ICARE/CATHARE computer codes, developed by IRSN (France) (the former with GRS, Germany), are used in RRC KI (Russia) for the analyses of accident transients on VVER-type NPPs. The latest versions of the codes were continuously improved and validated to provide a better understanding of the main processes during hypothetical severe accidents on VVERs.This paper describes modelling improvements for VVERs carried out recently in the ICARE common part of the above codes. These actions concern the important models of fuel rod cladding mechanical behaviour and oxidation in steam at high and very high temperatures. The existing models were improved basing on the experience in the field and latest literature data sources for Zr + 1%Nb material used for manufacture of VVERs fuel rod claddings.Best-fitted correlations for the Zr alloy oxidation through a broad temperature range were established, along with recommendations on model application in clad geometry and starvation conditions. A model for the creep velocity was chosen for the clad mechanical model and some cladding burst criteria were established as a function of temperature.After verification of modelling improvements on Separate Effect Tests, validation was carried out on integral bundle tests such as QUENCH, CODEX-CT, PARAMETER-SF (the application to the CORA-VVER experiments is not described in the present paper) and on the Paks-2 cleaning tank incident. The comparison of updated code results with experimental data demonstrated very good numerical predictions, which increases the level of code applicability to VVER-type materials.     

基于网格计数的停机剂量率计算方法研究

本文根据严格二步法的计算理论,研究了基于网格计数的停机剂量率计算方法,设计并实现了基于网格计数的停机剂量率计算程序。该程序能够支持圆柱坐标下的网格计算。本文使用源子程序进行复杂源描述。为了加快计算速度,本文采用了多节点和多线程等技术。本文利用国际热核聚变实验堆(ITER)停堆剂量基准实验ITER-T426进行测试,计算结果与实验值吻合良好,证明了该方法和程序的正确性和可用性。     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

蒙特卡罗粒子输运计算自动建模程序系统的研究与发展

在广泛调研和分析现有几何建模方法特点的基础上研发了具有可视化用户界面的自动建模程序系统MCAM.它可以实现多种商用软件CAD模型与MCNP模型之间的相互转换,且提供了模型建立、预处理、属性分析等基本功能和计算结果可视化及基于医学映像建模接口等扩展功能.全面系统地介绍了MCAM的设计思想与原理、总体结构、主要功能和国际合作协议框架下的应用测试等情况.实践表明,它是一个实用的MCNP计算辅助工具和核设计与核分析质量保证工具.