反应堆压力容器辐照监督

介绍了秦山第二核电厂600 MW压水堆机组的辐照监督计划,对监督、试验、评价方法以及超前因子偏大的原因进行了分析讨论.根据辐照监督数据评价了秦山第二核电厂反应堆压力容器辐照脆化效应.     

反应堆压力容器老化敏感性分析方法

结合近期开展的大亚湾反应堆压力容器老化分析及大纲编写工作,归纳总结了反应堆压力容器老化敏感性分析方法,提出了较为明确的表单化的老化分析流程,可以为相关的老化分析与评价活动提供借鉴.     

田湾核电站反应堆压力容器承压热冲击分析

反应堆压力容器(RPV)是核反应堆中不可替换的关键设备。田湾核电站在役前检查阶段,发现反应堆压力容器2#焊缝存在超标缺陷,2#焊缝处于堆芯筒体段,属强辐照区。为评价该缺陷的可接受性,采用有限元方法对反应堆压力容器2#焊缝进行了承压热冲击分析,在分析中考虑了小破口失水事故和安全阀误开启这两种最严酷工况。计算结果表明:有限元分析的结果与外国专家推荐方法的计算结果基本吻合,且田湾核电站反应堆压力容器2#焊缝寿期末的脆性转变温度小于最低容许脆性转变温度,能满足防脆断的设计要求。     

低铜合金反应堆压力容器钢辐照脆化预测评估模型

反应堆压力容器（RPV）材料辐照脆化预测评估对保证核反应堆安全运行、预防重大灾难性事故的发生具有重要意义。通过深入了解RPV材料辐照损伤机理和分析国外较为成熟的RPV辐照脆化预测模型,揭示了国外有关压力容器辐照脆化预测模型对低铜RPV辐照脆化预测的不足及其原因。在此基础上,发展和建立了适用于低铜RPV辐照脆化趋势的预测模型CIAE-2009。利用辐照性能数据对CIAE-2009模型进行了验证。结果表明,CIAE-2009对低铜含量RPV材料辐照脆化趋势预测具有较高的准确性和可靠性。     

大亚湾核电站2号机组反应堆压力容器老化现状的初步分析

为确保核电站设备在整个寿期内设计安全裕度要求能够得到满足,必须对设备老化进行有效的管理。对影响反应堆压力容器(RPV)的老化机理进行了初步分析,并结合大亚湾核电站的实际情况对2号机组RPV的目前状态进行了分析评估：     

核电厂反应堆压力容器老化管理大纲开发

根据IAEA系统化老化管理的理念和USNRC以执照更新为核心的老化管理方法出发,论述了核电厂反应堆压力容器老化管理大纲开发中需要考虑的要素.从法规体系、设备老化管理的基本要求、主要老化机理分析、文件体系审查及两种老化管理模式的适用性等角度,全面叙述了反应堆压力容器老化管理大纲开发中涉及的内容.以典型核电厂反应堆压力容器为例,给出老化管理大纲的工程应用实例.     

反应堆压力容器辐照脆化状态评估

基于大量相似辐照脆化试验测试数据和实际辐照监督测试数据,采用统计分析的方法,选出适合于某核电厂反应堆压力容器（RPV）的辐照脆化评估公式。以该核电厂已经完成的辐照监督管测试数据为输入,对RPV当前的辐照脆化状态进行了评估,并推算、分析了RPV在寿期末的结构完整性;基于辐照脆化计算结果,绘制了各运行阶段RPV的压力-温度限值曲线（P-T曲线）,并给出运行建议。      

AP1000反应堆压力容器承压热冲击下结构完整性分析

反应堆压力容器结构完整性是核电厂运行及延寿时需重点关注的问题之一。特别是承压热冲击（PTS）工况下反应堆压力容器结构完整性的验证工作对电厂能否安全运行有重要意义。为验证AP1000反应堆压力容器的结构完整性,本文简要阐述了AP1000反应堆压力容器进行确定性结构完整性分析的必要性,并对压力容器在典型PTS瞬态下的结构完整性进行了评价。分析评价采用概率断裂力学软件FAVOR中的FAVLoad模块进行,并应用IAEA-TECDOC-1627中的基准考题对该模块进行了验证,最后对AP1000反应堆压力容器进行了确定性结构完整性评价。评价结果表明,AP1000反应堆压力容器寿期末实际RTPTS值低于假想PTS瞬态对应的限值。反应堆压力容器在典型PTS瞬态下的结构完整性可以保证,同时也说明采用FAVLoad模块进行反应堆压力容器确定性结构完整性评价的方法可行。     

反应堆压力容器用Ni-Cr-Mo-V钢焊缝热老化脆化行为研究

对反应堆压力容器用Ni-Cr-Mo-V钢焊缝温度监督样品的热老化脆化行为进行了研究。焊缝属于压力容器的薄弱环节,服役时间最高达120 430 h（服役温度归一化到300 ℃）。3批次的焊缝监督样品冲击实验表明,焊缝材料在热老化过程中发生了脆化。通过研究发现,金相组织和显微维氏硬度在热老化期间未发生明显的变化,表明在热老化过程中不存在硬化脆化机制。断口分析及扫描俄歇纳米探针研究表明,晶界发生了P的偏析,弱化了晶界结合力,因此,反应堆压力容器用Ni-Cr-Mo-V钢焊缝在热老化过程中发生了由杂质元素P偏析引起的非硬化脆化。     

压水堆核电厂反应堆压力容器辐照脆化评价与监督

反应堆压力容器是压水堆核电厂的核心关键设备,受快中子(E1MeV)辐照造成的辐照脆化是其运行失效的重要因素,因此需要对压力容器进行辐照评价与监督,以保证其寿期内的安全运行。     

HEXTRAN–SMABRE calculation of the VVER-1000 coolant transient-1 benchmark

The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps are in operation. The problem is based on an experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, an extreme scenario concerning a control rod ejection after switching on a main coolant pump was calculated. At VTT the three-dimensional advanced nodal code HEXTRAN is used for the core dynamics, and the system code SMABRE as a thermal hydraulic model for the primary and secondary loop. The parallelly coupled HEXTRAN–SMABRE code has been in production use since early 1990s, and it has been extensively used for analyses of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used at VTT. The whole core calculation is performed with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation were specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Parametric studies have been performed for selected parameters.     

Modelling effects and their impact on the results of ATHLET solutions of Exercise 1 of the VVER-1000 Coolant Transient Benchmark

The VVER-1000 Coolant Transient Benchmark consists of two phases and refers to experimental data from the Kozloduy Unit 6 Nuclear Power Plant in Bulgaria. The paper describes the modelling features and their impact on the results of the Exercise 1, Phase 1 of the Benchmark obtained by two ATHLET user groups, namely GRS and NRI. The simulated transient is a main coolant pump (MCP) switching on in one loop at reduced power while three other MCPs are in operation. Particular attention is paid to the influence of the reactor vessel modelling and especially of the nodalization in the upper plenum. The comparison and discussion of the two simulation results confirm that the two solutions with the ATHLET system code achieve quite good system response of the plant transient.     

反应堆压力容器钢的辐照脆化预测模型研究

对反应堆压力容器（RPV）钢的辐照脆化进行预测是保证核电站长寿期安全运行的重要方法。通过深入分析国外已有RPV钢的辐照脆化预测模型,揭示了已有参数化预测模型的不足,在此基础上建立了新的预测模型PMIE-2012。利用辐照监督数据对PMIE-2012的可靠性进行评价,结果表明,PMIE-2012对RPV钢的辐照脆化预测具有较高的准确性和可靠性。     

基于压水堆堆芯物理分析软件Bamboo-C的VVER-1000工业确认研究

本研究旨在实现对VVER-1000型核反应堆的精确物理分析,基于西安交通大学自研的先进压水堆堆芯物理分析软件Bamboo-C,进行了深入的方法学研究。研究内容包括：在组件计算方面,研究了基于构造实体几何的六角形输运计算方法及重反射层的精细建模技术;在堆芯计算方面,研究了保角变换与非线性迭代策略结合的六角形节块中子扩散计算法。基于Bamboo-C软件对某VVER-1000机组连续3个燃料循环启动物理试验和功率运行进行了建模计算,并与实测数据进行了对比分析。结果表明：①启动物理试验中,临界硼浓度的误差均值为−5.0ppm（1ppm=10^–6）;慢化剂温度系数与等温温度系数的误差均值分别为0.3 pcm/K和0.9 pcm/K（1pcm=10⁻⁵）;硼微分价值的误差均值为−5.0%;控制棒价值的误差均值为−7.8%;②功率运行中,3个循环临界硼浓度的误差均值分别为−2.3ppm、−18.9ppm和−7.8ppm;3个循环的堆芯功率分布误差的均值为−0.010（组件相对功率大于1）和0.012（组件相对功率小于1）。因此,Bamboo-C软件对VVER-1000机组堆芯关键物理量的计算误差均满足工业限值要求,具备工程应用的能力。     

WER-1000型反应堆停堆及临界模式优化研究

在确保安全的前提下,经济性是核电厂的重要目标之一.VVER-1000型反应堆某些非并网运行的工况,如换料后重新临界、热停堆及临界、试验后返临界等操作,在操作所占用的时间、原材料的消耗量以及产生的废水量等方面可作优化.笔者对影响停堆及临界操作的重要因素,即控制棒和硼酸浓度的配置进行定性和定量的分析,得出优化的一般步骤和基本原则,并对3个案例实施了优化.     

辐照温度对反应堆压力容器材料辐照脆化行为的影响规律研究

反应堆压力容器（RPV）承受着强烈的中子辐照作用,随着快中子注量的累积,RPV产生不可忽视的辐照损伤,其中辐照温度是影响其辐照损伤的重要因素之一。针对辐照温度对RPV的影响机理研究,本文开展了现有预测模型分析、原位离子模拟辐照试验及多尺度模拟计算。对比分析了常用的辐照脆化预测公式,其温度适用范围为275~310℃,不适用于低温辐照条件。开展了不同温度下的原位离子辐照试验,结果表明辐照温度越高,辐照位错环尺寸越大而密度越低。多尺度模拟计算结果表明,辐照温度对辐照点缺陷的产生过程影响不明显,但对辐照缺陷的演化和平衡过程具有较明显的影响;辐照温度越低,材料辐照脆化越严重。研究揭示了辐照温度对RPV材料辐照脆化行为的影响机理及规律。     

Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

VVER-1000型机组正常运行气液态流出物源项计算

田湾3、4号机组采用俄方设计制造的VVER-1000型反应堆,其正常运行气液态流出物排放源项是检验核电厂设计是否满足国家相关环境标准的重要指标,是辐射防护最优化设计的重要内容之一。以我国压水堆核电厂源项框架体系为基础,通过分析田湾核电站相关工艺系统流程,选取合适的工艺回路部件数学模型,采用电厂设计以及实际运行经验参数,分别计算了设计与现实排放源项,并与俄方计算结果进行对比,说明采用新源项框架体系下气液态放射性流出物的变化情况。     

SHTV,as a technique for core calculation using spatial homogenization and temperature variation

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent homogeneous mixture of these material for determining the few group constants, while the effect of temperature variation in the fuel and coolant density along the axial core direction is considered. All calculations are performed using WIMS and CITATION codes. The obtained results are compared with the results of Final Safety Analysis Report (FSAR) prepared by the designer, and good agreement between the two results is shown.     

Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor

The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions.