截短型燃料组件堆内贮存临界安全分析

基于小型压水堆特有的截短型燃料组件,建立乏燃料贮存水池几何模型,分析正常贮存及事故工况下的临界安全。选取合理的保守假设,建立适当的计算模型,分别计算了一区和二区正常贮存工况、地震事故工况、组件跌落事故工况、新组件误插入事故工况的反应性。计算得到事故工况下有效增值因子最大值为0.932 83。小型压水堆乏燃料贮存水池临界安全分析中,正常工况及事故工况下计算结果均小于0.95。该设计模型可确保燃料堆内贮存区域处于次临界状态,且安全可控。     

高温气冷堆蒸汽发生器舱室混凝土温度场分析

采用ABAQUS6.7有限元分析软件,对高温气冷堆蒸汽发生器舱室混凝土在正常工况和设备冷却水系统停止供水事故工况下的温度场进行了计算。结果表明,在正常工况下,蒸汽发生器舱室混凝土的最高温度低于规定的限值;在设备冷却水系统停止对屏蔽冷却水系统供水事故工况下,7天内混凝土最高温度低于100℃,屏蔽冷却水系统能够保证对蒸汽发生器舱室的冷却。     

大型先进压水堆硼稀释事故模拟分析

基于美国先进三维节块法堆芯计算程序,建立大型先进压水堆堆芯首循环,选取四个最不利的保守事故工况点,包括满功率工况、启动工况、热备用工况、冷停堆工况,分别进行硼稀释事故分析,计算得到初始条件下的硼浓度以及硼稀释事故的临界硼浓度,最终计算总的硼稀释时间、报警发生时间以及从报警到临界的时间,分析大型先进压水堆发生硼稀释事故工况下的安全性。计算结果表明:在发生硼稀释事故工况下,反应堆有足够的时间在丧失全部停堆裕量前终止硼稀释。     

基于SCALE的RFA改进型燃料组件堆内贮存临界安全分析

RFA改进型燃料组件是西屋公司设计的能应用于大功率先进压水堆的改进型燃料组件。SCALE计算程序是一款在国际上得到广泛认可的综合性建模及模拟程序包,可用于核设计与核安全分析。基于SCALE计算程序,针对大功率先进压水堆的乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,分析乏燃料水池正常贮存及事故工况下的临界安全。计算结果表明一区正常贮存工况keff值为0.901 29,组件跌落事故工况下,有效增值因子为0.907 93。二区正常贮存工况下,计算模型keff值为0.909 98,新燃料组件误插入事故工况keff值为0.924 07。先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。     

中国实验快堆额定工况下冷热钠池数值分析

应用计算软件STAR-CD对中国实验快堆(CEFR)正常运行工况中的额定工况进行了三维数值分析，使用多孔介质模型对屏蔽柱的影响进行了模拟，给出了冷热钠池的三维温度场和流场，与已有热工设计进行了比较，并着重分析了浮升力在数值模拟计算中的影响，为事故工况下的设备动态分析及相应的设备力学分析提供了数据。研究结果为CEFR的优化设计及事故分析提供了参考数据和技术支持。     

X射线快速安全检查系统的辐射安全性分析

X射线快速安全检查系统是港口、物流中心、公路卡口等场所对大量进出集装箱货物进行快速安全检查的X射线类检查系统。国内某厂商研制的X射线快速安全检查系统, 采用司机直接驾车通过检查通道的方式来实现对集装箱卡车货物的快速检查, 本文介绍该系统的工作流程, 分析正常工况下司机被系统误扫、人员闯入或藏匿于集装箱中的安全风险, 以及设计基准事故下的安全风险。结果表明, 该系统是安全可行的, 关键人群司机以及意外闯入检查通道的人员所受剂量低于相关标准的要求, 对人员健康无影响。     

超功率下金属燃料钠冷快堆的动态仿真

编制了计算金属燃料钠冷快堆在超功率事故下的动态过程的仿真程序ＭＦＴＯＰ，并对它对美国池式钠冷快堆ＥＢＲ－Ⅱ在启动和功率运行工况下的反应引入事故瞬态进行了大量的分析计算，所得结果与国外大型程序ＮＡＴＤＥＭＯ的相应预测结果符合良好。本程序可用于其它钠冷快堆的超功率瞬态计算。     

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。     

《工业X射线探伤放射卫生防护标准》在环境影响评价中的几点探讨

固定式X射线探伤项目的环境影响评价中,职业卫生标准《工业X射线探伤放射卫生防护标准》(GBZ 117-2006)(以下简称GBZ 117-2006)为主要使用的技术标准与评价依据之一。针对固定式X射线探伤环境影响评价过程中遇到的实际问题,包括探伤房曝光室的屋顶、通风孔、电缆管口以及防护门等辐射防护薄弱环节,从实践性与可操作性方面展开讨论,对GBZ 117-2006提出完善与细化的建议。     

核电厂数字化主控室操纵员界面管理任务特征的研究

首先采用系统性人因失误减少和预测方法（SHERPA）分析操纵员界面管理任务中的关键行为；再用行为分析软件（INTERACT9）分析国内某数字化核电厂全范围模拟机上操纵员的操作录像视频，之后对INTERACT9采集的关键行为数据进行统计分析，得到4个操纵员界面管理任务的一般特征:①一、二回路操作员操作菜单栏、选择监视目标和打开参数界面的频率最高；②操纵员在选择进入不同系统界面的方式上趋向于选择从菜单栏进入；③一回路操纵员在正常工况和事故工况下的界面管理任务没有明显差异；二回路操纵员在正常工况下的界面管理任务明显少于事故工况；④正常工况下，一回路操纵员的界面管理任务显著多于二回路操纵员；事故工况下，一回路操纵员的界面管理任务与二回路操纵员的的界面管理任务相当。      

一起医用电子加速器辐射事故分析及救援概况

本文结合2001年波兰比亚韦斯托克肿瘤中心(BOC)医用电子加速器辐射事故及该机构的放射治疗设备概况,对波兰电离辐射安全监管体系进行了介绍,并对辐射事故过程、应急响应、IAEA救援、剂量评估、临床过程、结果和经验教训等方面进行了分析和说明。实践表明,导致向患者输出剂量率比预期高许多倍的原因包括:医疗机构的供电不稳定,NEPTUN10P型医用加速器不符合IEC颁布的最新标准,电子枪灯丝电流限制值设置在较高的水平,束流监测系统故障,二极管故障,安全联锁失效,显示屏剂量率低于实际值。IAEA援助小组的建议与援助、剂量评估以及良好的医疗条件为患者提供了医疗保障。本文可作为辐射事故应急的参考。     

严重事故下设备可用性论证要求

核动力厂应针对某些极不可能发生的严重事故进行设计已逐步成为共识,对在严重事故工况下需要保持安全功能的设备的质量要求也随之成为焦点问题,故进一步明确严重事故下设备质量要求及其验证方法和准则是落实核安全监管要求的重要组成部分。本文回顾了国内外关于核动力厂严重事故对策的发展历程,并解读了不同阶段对严重事故下所用设备的质量要求的内在含义。从我国相关核安全法规要求出发,结合我国核安全规划及远景目标,提出了严重事故下设备可用性论证的相关建议。     

Effects of accident management strategy on the severe accident environmental conditions

This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments.     

后处理厂溶剂再生系统始发事件的FMEA评价

识别始发事件是事故分析的基础。目前后处理厂对始发事件的识别尚未形成通用方法。本文以后处理厂共去污分离循环的溶剂再生系统为研究示范对象,采用失效模式和影响分析(FMEA)的工程评价方法识别和筛选始发事件。分析结果表明,该系统始发事件的类型主要包括:包容放射性物料的边界(设备、管道、阀门)破损泄漏;酸、碱洗槽界面测量仪表失效;各贮槽和洗涤槽液位测量仪表失效;污溶剂接受槽有机相出口计量泵轴封泄漏。经与美国后处理厂安全分析报告和国外后处理事故实例比较,FMEA方法分析结果对于设备失效所致的事故具有良好的包络性和适用性。因此,该方法可作为选取始发事件的参考方法,并可推广应用到后处理厂的其他工艺流程系统。     

核电厂主控室可居留性内漏剂量模型的研究

核电厂主控室无过滤渗漏风（内漏）的放射性影响是可居留性评价的重要部分,目前针对该部分的剂量模型过于简化,不符合实际工程设计。本研究结合核电厂实际设计特征,对内漏源项迁移机理进行研究,推导放射性活度微分方程,建立主控室可居留性内漏剂量模型,选取典型设计基准失水事故（LOCA）和发生堆熔的大破口失水事故（LB-LOCA）开展模型应用,并与目前常用的简化模型进行对比。结果表明,采用简化模型在LB-LOCA工况下的剂量结果小于采用本研究模型的结果,简化模型无法包络所有事故情景。经分析,本研究建立的内漏剂量模型更符合实际场景,适用于主控室可居留区域的内漏影响评价,并可用于内漏试验结果的验证以及工程项目设计。      

Ex-vessel core melt stabilization research (ECOSTAR)

The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical–chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO₂-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.     

核临界事故报警系统布置分析方法研究

本文阐述了布置核临界事故报警系统的意义和原则,分析了核临界事故可能发生的机理,初步建立了一套核临界事故情景假设分析方法。研究了最小临界事故源项计算方法以及三维剂量场分布计数的方法,采用各设备最小临界事故剂量场分布最小值等高线图的方法来从众多剂量场分布图中优化选取合适的核临界事故报警系统布置点位,以确保其可以覆盖到每个具有核临界事故风险的设备,并对核临界事故报警系统探头类型选择的原则和方法进行了分析。     

Comparison of UF6 packages under certification test and accident conditions

Abstract

Currently there are three packages approved by the NRC for US domestic shipments of fissile quantities of UF₆: NCI-21PF-1, UX-30, and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR part 71. The primary objective of this project was to compare conditions experienced during these tests to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR part 71 tests was achieved by means of computer modelling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from tests and other fire scenarios. In addition, the likelihood of encountering bodies of water during transport over representative truck routes was assessed. Modelled effects and their associated probabilities, accident rates, and other characteristics gathered from representative routes were combined with existing event tree data to derive generalized probabilities of encountering accident conditions comparable to or exceeding the 10 CFR part 71 test conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents.     

A Review of Nrc Instrumentation Needs

The accident at Three Mile Island highlighted the need to make improvements in nuclear power plant instrumentation. Since the accident, the Nuclear Regulatory Commission has required the installation of new equipment aimed at improving both post accident monitoring capability and the operator/equipment interface. Replacement of some existing equipment with qualified, highly reliable sensors and processing equipment will satisfy many of the new requirements. But implementation of other requirements will be more difficult. New measurement techniques must be developed or use of new technologies must be judged acceptable for safety purposes before some instrumentation needs can be satisfied. The lack of commercially available equipment has caused reassessment of the importance of other instrumentation needs. Looking beyond those modifications needed to correct deficiencies discovered as a result of the TMI experience, improvements in the reliability of normal plant operating equipment should be considered to reduce the frequency of safety system challenges. Advanced instrumentation presently available or under development may be useful in identifying equipment degradation, thereby preventing major equipment failures and consequent plant upsets. There is also a need for better diagnostic tools for improving the operator's response capability following plant upsets. Some of these tools are now available but others which show considerable promise require more development effort. Although significant improvements in safety may be possible through implementation of improved instrumentation, those improvements can be made only if the industry supports the development effort with both money and manpower.     

核电厂电仪设备延寿再鉴定试验研究

在核电厂电气仪表设备（简称电仪设备）环境鉴定研究成果的基础上,开展核电厂电仪设备延寿再鉴定分析和试验研究。以秦山第一核电厂DDG-1型电气贯穿件（EPA）为研究对象,根据运行实际制定了再鉴定试验研究的遵循原则,在此原则下结合分析法确定了试验方案和试验项目序列以及EPA修复依据和方案,并在此基础上开展再鉴定试验研究。适当修复后的DDG-1型EPA按试验大纲依次通过了设备性能随时间变化的试验、抗震试验、设计基准事故（DBA）条件下热力学试验和DBA后极限电性能试验,试验后状态完好,表明该DDG-1型EPA经适当修复后能够完成继续延寿20 a的预期目标,可为核电厂其他电仪设备再鉴定试验研究提供指导和借鉴。