压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

大型先进压水堆弹棒事故模拟及敏感性分析

相对于传统堆型,大型非能动先进压水堆堆芯设计具有重大改变,这些改变对弹棒事故分析具有重要影响,进而影响反应堆的安全性。通过选取典型的四类工况（寿期初满功率、寿期初零功率、寿期末满功率和寿期末零功率）,利用中子动力学软件和燃料性能分析程序开展大型先进压水堆CAP1400的弹棒事故模拟计算,验证大型先进压水堆弹棒事故工况下的安全性,并针对弹棒事故分析关键输入参数开展敏感性分析。计算分析结果表明：大型先进压水堆发生弹棒事故时,其结果能够满足验收准则的要求,反应堆处于安全可控状态;弹棒事故分析中功率峰值对弹棒价值最敏感,事故分析结果对停堆反应性敏感性较小。     

确定论数值反应堆程序NECP-X的开发及应用

数值反应堆是基于大规模并行计算平台,利用先进的物理模型和数值模拟算法,采用精细化建模,从而精确模拟反应堆在正常运行与事故工况中发生的各类物理现象的模拟技术。西安交通大学NECP团队基于自研的多群和连续能量数据库,提出了全局 局部耦合输运计算方法、大规模并行的2D/1D耦合输运方法等,开发了基于确定论方法的数值反应堆物理程序NECP X,并在此基础上实现了物理 热工 燃料性能分析的多物理耦合模拟计算。基于该程序及其耦合系统,在商用大型压水堆、研究堆和实验堆中进行了验证应用。数值结果表明,NECP X程序及其耦合系统可准确预测反应堆在运行过程中的关键安全参数随时间的演变情况,如有效增殖因数、功率、温度、应力、间隙宽度等,可为商用大型压水堆、研究堆和研究堆的设计及安全分析提供可靠的工具。     

CAP1400反应堆压力容器快中子注量独立审核计算

压力容器的使用期限:直接决定了反应堆的寿命,而快中子注量是影响其使用期限的重要参数之一,是核安全审评中关注的一项重要内容。作为核安全监管部门,对大型先进压水堆CAP1400的压力容器快中子注量进行审核计算,能够促进审评的独立性、科学性和有效性,为CAP1400的安全审评提供良好技术支持。本文利用蒙特卡罗方法分析程序对CAP1400反应堆压力容器快中子注量进行独立审核计算,并将计算结果与反应堆设计方利用离散纵标法所得结果进行对比。结果表明,CAP1400反应堆压力容器快中子注量审核计算结果与设计值的相对偏差在10%以内,并且快中子注量值满足标准审评大纲的相关要求。     

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。     

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。     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA FRTMB分析了CAP1400反应堆直接注射（DVI）管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA-FRTMB分析了CAP1400反应堆直接注射(DVI)管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA-FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

新书推荐:《大型先进非能动压水堆CAP1400》

正"大型先进非能动压水堆CAP1400"是在举国体制下开发的核电型号。"十三五"国家重点图书出版规划项目"核能与核技术出版工程"丛书之《大型先进非能动压水堆CAP1400》(上下册)一书从国家科技重大专项"大型先进压水堆CAP1400"的科研和工程实践出发,全面系统地介绍了CAP1400的总体技术、堆芯设计、系统设计、布置结构、试验研究、电厂运行、技术经济评价等内     

CAP1400反应堆堆芯入口流量分配试验研究

反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况（均匀流量工况、非均匀流量工况、偏回路流量工况）下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。     

AP1000机组首次临界试验验证计算

首次临界试验是压水堆核电厂调试启动过程的关键环节,旨在确认核反应堆堆芯能按照设计要求达到预期的临界运行状态。本文利用西安交通大学自主研发的NECP-Bamboo程序系统对AP1000机组堆芯的首次临界试验的设计结果进行了验证计算,并与AP1000堆芯的核设计结果进行了比较。计算结果表明：预估临界状态下的硼浓度的偏差为-15 ppm,控制棒积分价值的最大偏差为-52 pcm,硼微分价值的偏差不超过0.2 pcm/ppm,反应性温度系数的偏差不超过1 pcm/K。本文计算结果的精度与高保真计算程序KENO（概率论方法）和VERA（确定论方法）的计算精度相当,为确保AP1000堆芯调试启动阶段的核安全提供了进一步的数据支撑。     

IAEA GT-MHR benchmark calculations by using the HELIOS/MASTER physics analysis procedure and the MCNP Monte Carlo code

The IAEA's gas-cooled reactor program has coordinated international cooperation for an evaluation of a high temperature gas-cooled reactor's performance, which includes a validation of the physics analysis codes and the performance models for the proposed GT-MHR. This benchmark problem consists of the pin and block calculations and the reactor physics of the control rod worth for the GT-MHR with a weapon grade plutonium fuel. Benchmark analysis has been performed by using the HELIOS/MASTER deterministic code package and the MCNP Monte Carlo code. The deterministic code package adopts a conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation.In order to solve particular modeling issues in GT-MHR, recently developed technologies were utilized and new analysis procedure was devised. Double heterogeneity effect could be covered by using the reactivity-equivalent physical transformation (RPT) method. Strong core–reflector interaction could be resolved by applying an equivalence theory to the generation of the reflector cross sections. In order to accurately handle with very large control rods which are asymmetrically located in a fuel and a reflector block, the surface dependent discontinuity factors (SDFs) were considered in applying an equivalence theory. A new method has been devised to consider SDFs without any modification of the nodal solver in MASTER.All computational results of the HELIOS/MASTER code package were compared with those of MCNP. The multiplication factors of HELIOS for the pin cells are in very good agreement with those of MCNP to within a maximum error of 693 pcm Δρ. The maximum differences of the multiplication factors for the fuel blocks are about 457 pcm Δρ and the control rod worths of HELIOS are consistent with those of MCNP to within a maximum error of 3.09%. On considering a SDF in the core calculations, the maximum differences of the control rod worths are significantly decreased to be 7.7% from 21.5%. It is showed that there are good consistencies between the deterministic code package and the Monte Carlo code from the results of these benchmark calculations. Therefore, the HELIOS/MASTER 2-step procedure can be used as a standard reactor physics analysis tool for a prismatic VHTR.     

通用型中子输运程序VITAS应用研究

为提高确定论全堆芯中子输运程序的适用性,开发了通用型中子输运程序 VITAS。针对TAKEDA3 基准题(矩形组件)、TAKEDA4 基准题(六角形组件)、Dodds 基准题(R-Z 几何)和 C5G7-TD5 基准题(压水堆高保真计算)的验证结果表明,高阶的空间和角度基函数能够使结果稳定地向参考解渐进收敛,达到与多群蒙卡相当的计算精度水平。与参考解相比,TAKEDA3 基准题有效增殖系数(k_eff)偏差低于 60pcm(1pcm=10^-5),控制棒价值偏差为-3pcm,中子通量密度分布均方根(RMS)偏差为 1.03%;TAKEDA4 基准题 k_eff偏差低于 20pcm,控制棒价值偏差为 32pcm,中子通量密度分布 RMS 偏差为 0.70%;Dodds 基准题的功率最大偏差低于 1%;C5G7-TD5 基准题的功率偏差低于 0.9%。本文研究表明 VITAS 有望成为一套精确求解中子输运问题的通用型计算工具。     

Finite element spherical harmonics (PN) solutions of the three-dimensional Takeda benchmark problems

A set of multi-group eigenvalue (K_eff) benchmark problems in three-dimensional homogenised reactor core configurations have been solved using the deterministic finite element transport theory code EVENT and the Monte Carlo code MCNP4C. The principal aim of this work is to qualify numerical methods and algorithms implemented in EVENT. The benchmark problems were compiled and published by the Nuclear Data Agency (OECD/NEACRP) and represent three-dimensional realistic reactor cores which provide a framework in which computer codes employing different numerical methods can be tested. This is an important step that ought to be taken (in our view) before any code system can be confidently applied to sensitive problems in nuclear criticality and reactor core calculations. This paper presents EVENT diffusion theory (P₁) approximation to the neutron transport equation and spherical harmonics transport theory solutions (P₃–P₉) to three benchmark problems with comparison against the widely used and accepted Monte Carlo code MCNP4C. In most cases, discrete ordinates transport theory (S_N) solutions which are already available and published have also been presented. The effective multiplication factors (K_eff) obtained from transport theory EVENT calculations using an adequate spatial mesh and spherical harmonics approximation to represent the angular flux for all benchmark problems have been estimated within 0.1% (100 pcm) of the MCNP4C predictions. All EVENT predictions were within the three standard deviation uncertainty of the MCNP4C predictions. Regionwise and pointwise multi-group neutron scalar fluxes have also been calculated using the EVENT code and compared against MCNP4C predictions with satisfactory agreements. As a result of this study, it is shown that multi-group reactor core/criticality problems can be accurately solved using the three-dimensional deterministic finite element spherical harmonics code EVENT.     

双面慢化环形燃料反应堆物理实验

环形燃料零功率反应堆是首个双面慢化环形燃料作为核燃料的反应堆。本文采用周期法、落棒法获取环形燃料零功率反应堆的临界参数、控制棒价值、元件价值、含Gd元件的反应性效应等关键参数,对环形燃料零功率反应堆的物理性能进行实验研究,验证环形燃料反应堆堆芯物理设计计算程序。结果表明：根据外推过程确定堆芯临界装载环形燃料元件96根,实心燃料元件172根,此时k_eff为1.000 40,堆芯调节棒价值为-247.5 pcm,安全棒价值为-1 358.4 pcm;元件价值与理论值平均偏差为1.3 pcm,含Gd元件反应性效应与理论值平均相对偏差为8.8%。本文结果为环形燃料的工程化设计程序提供关键数据支撑。     

Control rod worth calculations using deterministic and stochastic methods

Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results’ reliability.     

SARAX-FXS程序针对非均匀快堆堆芯计算中的改进及应用

SARAX-FXS程序是基于确定论方法,适用于快谱堆芯组件能谱、均匀化参数计算的程序。由于快堆中组件空间自屏的非均匀效应不可忽视,本文将基于一维圆柱、平板几何的碰撞概率方法加入SARAX-FXS模块,并以等效一维模型计算组件的均匀化参数。为保证能群归并前后的核反应率守恒,在组件计算中引入超级均匀化(SPH)因子修正截面。采用快堆基准题MET-1000对程序的计算结果进行验证,结果表明,与参考解相比,SARAX-FXS的一维计算模块具有较高的精度,特征值计算相对偏差在100~200pcm之间。堆芯计算结果显示,引入SPH因子可提高特征值计算的精度约300pcm,功率分布的均方根误差可从约3%下降至约1%。     

压水堆重反射层堆芯核热耦合高精度计算分析研究

重反射层的应用可提高反应堆中子经济性,其结构和中子吸收特性均与压水堆常规围板/反射层差异较大,因此对核设计程序的计算分析能力提出了新的要求。为分析重反射层建模方案对堆芯中子学计算结果的影响,使用先进中子学程序SCAP N和确定论堆芯高保真模拟程序NECP X对压水堆重反射层问题进行了高保真模拟,分析了5种反射层建模方案下计算结果的差异,并将高精度计算结果与商用核设计程序系统进行了对比。数值结果表明,重反射层水洞内冷却剂温度变化对计算结果影响较小;相较精确建模方案,重反射层铁水打混建模方案造成的反应性计算偏差在±30 pcm以内、组件相对功率分布计算偏差在±2%以内。     

球床高温气冷堆吸收体价值计算的验证

VSOP程序广泛用于球床高温气冷堆的工程设计。对于被布置在堆芯侧反射层孔道中、用于反应性控制的吸收体,由于物理计算方法的限制,VSOP程序不具备计算其价值的功能,必须借助其他确定论程序进行外部耦合计算,涉及到几何的近似处理、截面的归并和转换,可能引入额外的误差。为此,本文采用蒙特卡罗程序建立了精细的堆芯模型,真实描述了堆芯活性区的球床结构、侧反射层的孔道结构、吸收体的形状和位置,在同样的堆芯状态下,比较了确定论耦合程序和MCNP程序计算得到的吸收体价值。结果表明：确定论耦合程序的计算结果是准确的,从设计角度上是偏保守的。