Non LOCA-三维物理GINKGO/COCO?耦合程序的开发与验证

针对更为精细和准确的堆芯建模与热工水力分析需求,基于自主研发的Non LOCA热工水力分析程序GINKGO和三维物理程序COCO,采用动态链接库（DLL）技术开发了GINKGO/COCO耦合程序;介绍了耦合程序的开发原理和实现方式,并采用经济合作与发展组织（OECD）主蒸汽管道破裂事故（MSLB）国际基准题对其进行了验证。结果表明,GINKGO/COCO耦合程序的计算结果与OECD MSLB国际基准题的结果较为吻合。因此,GINKGO/COCO耦合程序具有良好的计算能力和可靠性。      

三维物理-热工耦合系统RECON的开发与验证

基于RELAP5、COBRA-Ⅳ和NLSANMT程序,采用并行耦合模式与并行虚拟机技术,开发了三维物理-热工耦合系统RECON,其耦合形式灵活,可根据分析需要选择用于耦合的程序。利用系列基准题进行了验证,特别是针对MSLB基准题的计算,与国际上众多耦合程序相比,RECON具有较好的计算精度,可用于反应性引入事故分析     

子通道程序SUBSC的开发及稳态验证

采用面向对象模块化编程技术开发了面向大规模热工水力计算的自主化子通道程序SUBSC,利用SUBSC和COBRA程序分别计算了典型压水堆1/4组件,结果表明,两者计算结果吻合很好。为进一步验证SUBSC程序,计算了PSBT稳态5×5棒束基准题,结果表明,在各种工况下SUBSC程序计算得到的通道平均含汽率与实验测量值吻合很好,最大相对偏差仅为0.7%,证明了程序具有较高的计算精度。为提高SUBSC程序的计算效率,引入不完全LU分解预处理的再启动GMRES算法求解质量守恒方程,对多组件的计算结果表明,SUBSC程序具备大规模热工水力计算能力。     

用OECD/NEA堆芯瞬态基准题验证RELAP5-TDNK程序系统

利用经济发展与合作组织核能机构(OECD/NEA)压水堆堆芯弹棒瞬态基准题对RELAP5-TDNK进行了验证.使用RELAP5-TDNK建立了弹棒基准题模型,分析了两种弹棒问题,对程序的数据交换能力、耦合方法和瞬态事故分析能力进行了检验.与国际上多种程序进行比较,结果表明:RELAP5-TDNK程序模拟结果较好,能够分析事故或瞬态过程中堆芯内局部功率和热工参数的相互作用,具有分析强反馈现象的能力.     

稳态堆芯多物理耦合系统CSSS V1.0的研发

充分考虑反应堆堆芯中子学物理、热工水力、燃料等专业的相互耦合过程,将先进节块法堆芯中子学计算软件NACK V1.0、热工水力子通道软件CORTH V2.0、燃料棒性能分析软件FUPAC V1.1进行集成耦合,得到稳态堆芯多物理耦合模拟设计分析系统CSSS V1.0,可计算典型压水堆的稳态运行物理、热工、燃料等专业参数。通过NEACRP-L-335压水堆基准问题验证计算,CSSS V1.0系统的计算结果与国际基准PARCS程序总体符合较好。     

三维耦合分析软件的落棒事故分析

作为数值反应堆中必不可少的物理和热工部分,中广核研究院有限公司开发了三维物理热工耦合分析软件,通过动态链接库技术实现了自主研发的核反应堆系统瞬态分析软件和三维核设计软件的耦合,并已与国际基准题结果对比验证。本文为耦合软件的应用,围绕华龙一号的落棒分析问题,开展不同落棒组合的耦合计算分析,并研究停堆棒组落棒和温度调节棒（R）棒组两组落棒对堆芯功率的影响。分析结果表明,非中心对称的棒组落棒事故会导致堆芯径向功率出现不对称,并使得堆芯出口回路温度不同。落棒反应性价值越大,R棒调节后的稳态功率回升相比初始稳态差异越大,DNBR公式计算值的变化趋势与功率呈现相反规律。     

基于格林函数节块法的物理与热工-水力耦合方法研究

开发了三维物理与热工-水力耦合的PWR堆芯瞬态分析程序NGFMN-K/COBRA-Ⅳ/COBRA-Ⅳ(NCC)。少群时空中子动力学计算采用格林函数节块法程序NGFMN-K,隐式耦合子通道程序COBRA-Ⅳ实现瞬态计算。采用P10H8B功率重构方法给出热组件栅元功率分布,耦合另一个COBRA-Ⅳ程序模块,进行热组件子通道分析得到安全参数。对NEACRP-L-335 C1弹棒基准问题的计算表明,NCC程序的计算结果与参考结果符合很好,说明程序计算正确,可用于评估事故结果。     

小型铅-铋冷却快堆提棒事故核热耦合研究

基于热工程序COBRA-YT和物理程序SKRTCH-N,利用并行虚拟机（PVM）平台开发了核热耦合工具:COBRA-YT将冷却剂密度和燃料温度等热工参数传递给物理程序,用以更新截面;SKETCH-N执行物理计算,并将功率分布反馈给热工程序;最后,应用该耦合程序分析铅-铋冷却快堆的提棒事故。计算结果显示控制棒提起后,功率迅速升高,在1.42?s后达到最大值;5?s后包壳温度达到峰值1264℃,超出了设计限值。结果表明:在提棒事故后,均一化布置堆芯的安全会在极短时间内受到严重威胁,故该堆芯应采用分区布置。      

基于高阶CRAM的燃耗程序开发与验证

本文基于高阶切比雪夫有理近似方法（CRAM）研制了点燃耗程序ICRAM,并内耦合于蒙特卡罗输运程序OpenMC,形成了一套燃耗计算分析程序OPICE。与传统部分分式分解（PFD）形式的CRAM相比,高阶不完全局部分解（IPF）形式的CRAM具有数值稳定性好、计算精度高和步长包容性更好等特点,满足高保真燃耗计算发展的需求。为提高耦合计算精度,OPICE采用了预估-校正和子步法两种耦合策略,支持纯衰变、定通量和定功率3种计算模式。通过OECD/NEA压水堆栅元燃耗基准题和快堆燃耗基准题的验证,程序计算结果与实验值及各参考值吻合良好,初步验证了OPICE的正确性与有效性。     

基于统一框架的多物理耦合方案研究与平台开发

为实现反应堆多物理、多过程、高保真数值计算,捕捉堆芯内部更真实的物理学行为,本文深入研究了多物理程序耦合方案,并基于上层监控架构、串行计算模式、网格一一映射的显式耦合方案,依托开源集成平台SALOME、通用平台接口ICoCo、三维堆芯中子学程序ADPRES和系统热工水力程序RELAP5搭建了基于统一框架的多物理耦合平台。经NEACRP-L-335压水堆弹棒基准题验证表明,耦合平台计算结果与基准例题吻合良好,耦合平台在功率峰捕获上更加准确,可释放部分安全裕量;对瞬态末各参数的计算结果也有足够高的精度,证明了耦合平台可对反应堆多物理、多过程耦合工况进行更精细、更深入的数值计算与安全分析。      

Development and testing of the code for automatic generating of multi-temperature continuous-energy neutron cross section libraries

In the nuclear reactor design, a code for automatically generated multi-temperature continuous-energy neu- tron cross section data library, which is called AMTND for short, was designed and developed to meet the need of the reactor core design coupled with thermal-hydraulic design. The code can provide a point-wise cross- section at any temperature for a Monte Carlo neutron transport program, such as MCNE In ensuring that the nuclear data produced by AMTND meets the testing of critical benchmark experiments, the time-consumed by the nuclear data generating of AMTND compared with NJOY＇s was carried out and the result shows the code＇s excellence. In order to test the accuracy of the code, out and the results verified the code preliminarily. the Doppler coefficient test benchmark was also carried     

PARCS稳态两相热工水力程序PATHS的开发与耦合

本工作开发了PARCS的先进热工水力求解器PATHS,可对沸水堆进行热工水力稳态模拟。与RELAP5的计算结果进行验证,结果表明,PATHS的计算结果与RELAP5的基本一致。将PATHS与PARCS进行耦合,对SMART反应堆及Peach Bottom 2 OECD Turbine Trip基准题进行计算,结果表明,PARCS/PATHS耦合程序计算结果准确有效,能用于沸水堆的稳态物理热工耦合计算。     

SN2D程序在C5G7基准例题上的验证与分析

C5G7基准例题因其具有强烈非均匀性、组件能谱差异大等特点,常用来检验软件计算有效性。在对栅元进行不同的几何等效和对原有的栅元等效均匀化方法进行改进的情况下,采用二维离散纵标法输运计算程序SN2D对C5G7基准例题进行了分析计算,给出了各种计算条件下k_eff和功率分布的计算误差。结果表明,SN2D程序可应用于C5G7基准例题的求解。计算结果可为求解类似问题时计算程序及条件的选择提供直接参考。     

Multi-physics steady-state analysis of OECD/NEA modular high temperature gas-cooled reactor MHTGR-350

This paper presents the application results of MCS/GAMMA+ to multi-physics analysis of OECD/NEA modular high temperature gas-cooled reactor (MHTGR) benchmark Phase I Exercise 3. It is a part of international R&D efforts lead by the Next Generation Nuclear Plant (NGNP) US project to improve the neutron-physics and thermal-fluid simulation of (high temperature gas-cooled reactors) HTGRs, one of the next generations of safer nuclear reactors. Accurate and validated analysis tools are indeed a crucial requirement for safety analysis and licensing of nuclear reactors. To guide this effort, a numerical benchmark on the MHTGR was created by the NGNP project and formally approved in 2012 for international participation by the OECD/NEA. The benchmark defines a common set of exercises and the comparison of solutions obtained with different analysis tools is expected to improve the understanding of simulation methods for HTGRs. The coupled neutronics/thermal-fluid solution presented in this paper was obtained with the neutron transport Monte Carlo code MCS developed by Ulsan National Institute of Science and Technology and the thermal-fluid code GAMMA+ developed by Korean Atomic Energy Research Institute. The purpose of this paper is to present the GAMMA+/MCS coupled system, the calculation methodology, and the obtained solutions.     

The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems

In support of the pebble bed modular reactor (PBMR) Verification and Validation (V&V) effort, a set of benchmark test problems has been defined that focus on coupled core neutronics and thermal-hydraulic code-to-code comparisons. The motivation is not only to test the existing methods or codes available for high-temperature gas-cooled reactors (HTGRs), but also to serve as a basis for the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for design and safety evaluations in future.The reference design for the PBMR268 benchmark problem is derived from the 268 MW PBMR design with a dynamic central column containing only graphite spheres. Several simplifications were made to the design in order to limit the need for any further approximations when defining code models. During this process, care was taken to ensure that all the important characteristics of the reactor design were preserved. The definition and initial phases of the benchmark were performed under a cooperative research project between NRG, Penn State University (PSU) and PBMR (Pty) Ltd. However, participation has been extended to include Purdue University and INL. All contributions to the benchmark effort were made in-kind by the participating members including the participation in four benchmark meetings over a period of 3 years. Based on the work performed in this benchmark the PBMR 400 MW design with fixed central reflector has been accepted as an OECD benchmark problem and work has already started.In this paper, the benchmark definition and the different test cases are described in some detail. Phase 1 focuses on steady-state conditions with the purpose of quantifying differences between code systems, models and basic data. It also serves as the basis to establish a common starting condition for the transient cases. In Phase 2, the focus is on performing coupled kinetics/core thermal-hydraulics test problems with a common cross-section and material property sets. The six events selected are described, and examples of some results are included to illustrate the behaviour of the transients. The final results of this work will be published in an NRG report and the focus will move to the OECD 400 MW benchmark problem.     

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.

The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.     

基于最佳一致逼近多项式的燃耗计算方法研究

本文研究了一种基于最佳一致逼近多项式（MMPA）的燃耗计算方法求解燃耗方程。相比于切比雪夫有理近似方法（CRAM）和围道积分有理近似方法（QRAM）,MMPA方法只需一次矩阵求逆计算即可求解燃耗方程,且所有计算都是实数运算,具有数值稳定性好、求解效率高等优点。进一步研制了基于MMPA方法的点燃耗程序AMAC,并耦合蒙特卡罗输运程序OpenMC,采用衰变例题、固定辐照例题、OECD/NEA压水堆栅元燃耗基准题和沸水堆组件燃耗基准题进行验证,程序计算结果与实验值及各参考值吻合良好,初步验证了MMPA方法在理论和数值上的正确性和有效性。     

Development and Validation of Burnup Code Based on High-order Chebyshev Rational Approximation Method

Based on high-order Chebyshev rational approximation method (CRAM), a point-burnup code named ICRAM was developed and internally coupled to Monte Carlo code OpenMC, forming a burnup calculation and analysis program OPICE. Compared with the traditional partial fraction decomposition (PFD) form of CRAM, the high-order incomplete partial fractions (IPF) form of CRAM has the characteristics of good numerical stability, high calculation accuracy and better step tolerance, etc., which meets the needs of high-fidelity burnup calculation development. In order to improve the accuracy of coupling calculations, two coupling strategies including prediction-correction method and sub-step method were implemented in OPICE. Three different calculation modes were supported by OPICE to execute the decay, constant flux and constant power calculations. By calculating the OECD/NEA burnup benchmark and fast reactor burnup benchmark, the calculation results of OPICE are in good agreement with the experimental data and each reference value. The correctness and validity of OPICE are verified preliminarily.