高通量工程试验堆燃料元件热工水力特性计算

从基本的质量、动量、能量守恒方程出发,建立了合理的高通量工程试验堆多层套管元件的热工水力特性分析计算模型,并运用在此模型基础上开发的计算程序对高通量工程试验堆燃料元件的运行工况进行了分析计算,计算结果与理论分析以及高通量工程试验堆实际运行结果相符.     

反应堆输入线性正反应性时考虑温度反馈的仿真计算

运用点堆中子动力学方程建立模型,计算了某小型反应堆在6种典型工况下,反应堆引入线性正反应性时各主要参数的变化,并将计算结果与该堆的三维实时仿真软件的计算结果进行了比较.结果表明,点堆模型可以模拟出反应堆受到线性正反应性扰动后各主要参数的峰值和扰动后的稳定值,但在响应时间和波动持续时间方面仍需改进.     

应用MCNP程序处理研究堆屏蔽计算深穿透问题的方法研究

采用MCNP程序对某一体化研究堆主屏蔽设计进行验证计算。通过几何分裂、源项模型简化、多群截面和能群截断等方法逐步优化计算模型,计算效率提高了70%,计算结果方差小于10%,与确定论程序计算结果吻合较好。经分析,本文建立的计算方法适用于屏蔽层较厚研究堆的屏蔽计算。     

用于大型轻水堆三维瞬态分析的简化籍贯方程节块法

本文讨论了用于大型轻水堆三维瞬态分析的简化籍贯方程节块法,提出了把这种节块三维瞬态计算和点动力学计算耦合起来的近似模型,并用轻水堆三维动力学试验问题的数值计算结果同节块格林函数法和粗网通量展开法进行了比较。     

溶液堆物理计算程序FMCAHR燃耗功能及其验证

介绍了FMCAHR程序的燃耗计算模型及流程,并使用燃耗基准题和DRAGON程序对燃耗计算结果进行验证.验证结果表明,FMCAHR燃耗计算功能的准确性较高,适用于溶液堆的燃耗计算分析.     

AP1000堆芯仿真计算建模与精度分析

针对核电厂AP1000堆芯描述,建立由组件计算、截面拟合处理计算模型,并得到组件少群常数;采用两群三维,实时中子动力学仿真模型,选取11组衰变功率计算堆芯衰变功率的三维变化,同时为了准确计算反应堆的"中毒"变化,三维空间上考虑氙、钐以及先驱核碘、钜元素浓度的影响特性,建立针对AP1000堆芯实时仿真计算模型,并准确计算反应堆的"中毒"和氙振荡现象,为验证模型建立的正确性与堆芯实时仿真程序SimCore的精准性,对堆芯临界硼浓度、堆芯温度、控制棒价值进行计算,同时选取汽机停机不停堆、反应堆满功率跳堆运行,反应堆正常停堆运行及控制棒落棒、弹棒事故响应等不同测试工况,对结果进行验证及分析。结果表明:建立的三维堆芯实时仿真程序模具有较好的精准性,可以用于全范围模拟机堆芯计算,并广泛应用于核电厂堆芯物理仿真。     

脉冲堆失水事故计算程序的开发和应用

由于西安脉冲堆的特点，致使国际上通用的瞬时堆芯裸露模型不能使用。中国核动力研究设计院建立了反映西安脉冲堆失水事故机理和过程的真实真芯裸露模型，开发了相应的计算机程序，用于分析和评价西安脉冲堆的安全特性。分析结果表明，真实堆芯裸露模型具有广泛的实用性，可用于计算全部侧面破口和底部破口的失水事故。在破口直径相同的条件下，西安脉冲堆侧面破口失水事故后果比底部破口失水事故严重。在目前的设计条件下，即使发生失水事故，西安脉冲堆也能满足安全准则的要求。     

西安脉冲堆计算值和实测结果分析

介绍了西安脉冲堆堆芯布置、计算模型和程序以及计算值与实验测量结果分析。西安脉冲堆在满功率（2MW）下连续运行了72小时，证明西安脉冲堆整个系统和设备是良好的、安全可靠的。根据西安脉冲堆计算值与测量值的比较，证明计算程序可靠，具有较高精度。西安脉冲堆理论计算值与实验测量值符合较好，完全满足工程设计要求，实验测量结果达到了设计指标的要求。     

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。     

熔融物堆内滞留条件下压力容器外部自然循环特征分析

基于一维稳态两相分离流模型,并引入低流速过冷沸腾净蒸汽产生点方法,建立熔融物堆内滞留(IVR)条件下的两相流动数值计算模型,以获取两相流真实含气率。通过对比ULPU-V试验中自然循环流量,验证数值计算模型预测结果的准确性。针对AP1000堆型,分析几何结构参数和热工参数对其IVR策略中自然循环流量的影响。分析结果显示,冷却水过冷度、流道间隙、堆腔淹没水位、流道入口面积和出口阻力系数对自然循环过程有着不同程度的影响,自然循环稳态流量呈现出不同的变化趋势。     

Validation of coupled neutron kinetic/thermal-hydraulic codes. Part 2: Analysis of a VVER-440 transient (Loviisa-1)

Several three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal-hydraulic system codes. Under the auspices of the European Union's Phare programme these codes have been validated against real plant transients by the participants from 7 countries. Two of the collected five transients were chosen for validation of the codes. Part 1 of this article consists of validation against VVER-1000 reactor data. This second part is focussed to validation against measured data of ‘One turbo-generator load drop experiment' at the Loviisa-1 VVER-440 reactor. The experiment was performed just after plant modernisation and more measured data was available to validation than in normal operation of real plants. Good accuracy of the results was generally achieved comparable to the measurement accuracy. The confidence in the results of the different code systems has increased, and consequences of certain model changes could be evaluated.     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

六角形组件源项计算方法研究

为了满足水-水高能反应堆（VVER）或快堆屏蔽计算需求,编写了离散坐标（SN）和蒙特卡罗（MC）六角形组件源项计算程序,并且在VVER堆型上进行了验证。数值结果表明,对于吊篮内表面到压力容器外表面径向的三群中子和光子注量率分布,SN和MC计算得到的大部分中子和光子计算偏差小于30%,证明理论模型和程序是正确的,可以用于VVER或快堆的屏蔽计 ?      

The reactor dynamics code DYN3D – models,validation and applications

The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given.     

秦山一期堆本体退役源项估算及辐射场可视化

针对我国秦山一期核反应堆实际情况,利用蒙特卡罗程序建立了细化到燃料棒结构的全堆芯pinby-pin模型进行中子输运计算,并对计算模型的可靠性进行了验证;基于堆本体结构部件的几何参数、材料参数及堆本体中子注量率分布,在假定功率运行史的情况下,利用燃耗计算程序计算了反应堆停堆后的中子活化产物作为堆本体退役源项的估算结果,并对源项产生的三维辐射场剂量分布情况进行了可视化建模与分析,模拟结果与理论分析一致。本研究是下一步建立我国秦山核电厂退役技术安全验证和虚拟仿真平台的关键性基础工作。     

核电站反应堆辐射屏蔽程序系统

核电站反应截辐射屏蔽程序系统包括源项程序、离散座标输运程度、蒙特卡罗和反照蒙特卡罗程序、点核积分程序、最佳化程度、温场程序、大气散射和结构壁屏蔽效应分析程序、数据库以及加工程序和耦合程序，本程序系统程序类型比较齐全，程序和参数配大，在核电站反应堆以及其它类型反应堆和核设施辐射屏蔽设计和安全分析中得到了广泛应用。     

退役反应堆放射性活化源项计算

本文建立了退役反应堆活化源项的计算模型,通过临界计算验证了模型的正确性。介绍了对于距堆芯较远的区域采用分层计算和分步计算的重要意义,通过MCNP和ORIGEN程序相结合,计算了距堆芯较远处的支撑裙、铅支撑筒内侧和外侧钢板样品和一次水箱外筒的预留样品⁶⁰Co比活度,计算值与测量值的偏差满足退役工程设计需求,表明本文所建立的退役反应堆放射性活化源项计算方法和模型是适用的。     

Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.     

脉冲堆物理设计分析

本文介绍了我国自行设计建造的首座脉冲型实验反应堆的堆芯核设计、计算模型和程序以及计算结果与零功率实验值的分析比较。     

CFD在核能系统分析中应用的最新进展

介绍了国内外计算流体力学(CFD)方法在核能系统分析中应用的最新进展.CFD已经可以应用到一些三维单相瞬态流动情况中,其中包括堆芯及组件内的流场模拟,以及堆芯外空腔和其他领域的模拟分析.CFD的应用还需要进行进一步验证和基准化,并需要针对CFD的应用方法进行研究,包括建立最佳实践导则(BPG).CFD程序还被用于与热工水力系统程序的多尺度耦合以及与中子物理程序的耦合.CFD程序在两相流领域的应用还处于初级阶段,在湍流和多相流的模拟中需要更进一步发展.