仪控电子设备电磁兼容鉴定试验法规标准适用性分析

介绍了做安全级仪控电子设备电磁兼容（EM C ）鉴定试验时如何选择适用的法规标准。重点介绍了电磁干扰试验、射频辐射干扰试验、敏感性试验、电源浪涌试验以及静电放电试验等电磁兼容测试试验项目,并介绍了电磁兼容试验样机配置的重要性及选择方法。     

放射性物质运输货包安全试验

介绍了中国放射性物质运输遵守的法规和中国辐射防护研究院用于放射性物质运输货包试验的下落试验设施、耐热试验设施和数据获取能力。试验设施根据IAEA的《放射性物质安全运输条例》（TS-R-1）和中国的《放射性物质安全运输规程》（GB 11806-2004）的要求建设。下落试验设施能用于13 t级以下的A型和B型货包的自由下落试验、贯穿试验、力学试验（自由下落试验Ⅰ、自由下落试验Ⅱ和自由下落试验Ⅲ）。耐热试验设施能完成B型货包的耐热试验。利用这些设施已进行了FCo70-YQ型货包、30A-HB-01型货包、SY-I型货包和XAYT-I型货包的遵章取证试验     

反应堆堆内构件动态特性试验研究分析

介绍了反应堆试验模型主要部件在空气中和静水中进行的动态特性试验,包括所用的测量系统、试验原理、试验方法和得到的试验结果。并对试验结果进行了分析,说明了试验结果的合理性。     

反应堆物理试验用便携式数字反应性仪实堆考验试验

反应堆物理试验用便携式数字反应性仪的研制过程中,必须在零功率堆上进行实堆考验试验,以检验其测量功能及测量精度。叙述了实堆考验的试验装置、试验堆芯、试验方法、试验内容以及试验结果,试验结果表明:该仪器满足技术指标的要求,满足反应堆物理试验的反应性测量要求。     

核动力反应堆零功率物理试验的多普勒发热点测量试验

为了定量确定零功率物理试验功率的上限,在反应堆零功率物理试验中,利用数字反应性仪测量多普勒发热点,以确定试验功率的范围、保证试验精度。叙述了本次多普勒发热点测量试验的原理、试验仪器、试验方法、试验结果及数据处理方法等,试验结果表明:利用数字反应性仪测得的反应性经过修正后可以准确地判断多普勒发热点,可为后续物理试验提供参考。     

核一级电动截止阀样机试验

介绍核一级电动截止阀样机鉴定试验目的、试验内容、试验装置、测量参数、试验方法等内容。试验结果表明:试验前阀门阀座、阀杆密封无泄漏,冷态开启、关闭行程时间均为11 s;在热态寿命、热态循环、冷热交变及流体阻断试验过程中阀门开启、关闭正常,开启、关闭行程时间均为11 s;试验后阀门阀座密封10 min泄漏12滴水,阀杆密封无泄漏,冷态开启、关闭行程时间均为11 s,各项试验数据表明阀门满足设计要求。     

秦山第二核电厂物理试验管理

核电厂物理试验是验证反应堆核设计安全性的重要项目之一。秦山第二核电厂通过多堆年对物理试验管理的持续改进,制定了物理试验监督要求;结合试验文件,组织机构控制,试验流程控制,试验人员、设备、方法等各方面的管理,形成了一套较为完善的物理试验管理体系;为物理试验的安全、质量和时间控制提供了保障。     

核电厂非能动氢气复合器消氢特性试验研究

根据非能动氢气复合器(PAR)的工作状态特点和启动阈值、停止阈值、消氢能力、点火阈值等关键特性参数的要求,设计建立能够模拟安全壳内事故环境条件、在非能动条件下开展PAR关键特性参数验证试验的试验装置,制定相应的试验方法,开展启动阈值试验、启动时间试验、消氢能力试验和点火阈值试验等,获得PAR的关键特性参数。试验结果表明:PAR关键特性在不同的试验参数条件下测试结果也不同;在制定PAR消氢特性参数要求时需要限定试验方法和试验参数条件,以便获得统一的、定量的PAR的消氢特性参数。     

EPR核电机组部分冷却试验研究与风险识别

部分冷却试验作为欧洲先进压水堆（EPR）核电机组的首堆试验,试验过程中将造成压水堆一、二回路巨大的热冲击,核电厂全寿期内允许次数为15次。为减少重大瞬态试验的风险,本文对部分冷却试验的控制逻辑、试验原理及可行性进行深入研究,总结提炼出5大风险点。并通过模拟仿真对部分冷却试验进行分析,优化试验方案,最终试验一次成功满足核安全要求。      

秦山核电厂反应堆安全壳整体密封性能试验

简要介绍秦山核电厂反应堆安全壳的整体密封性能试验,包括试验方法、试验程序及过程、使用的仪表和测量系统;以及试验结果。文中还初步探讨了影响试验结果的一些重要因素。     

核电厂安全壳首次整体在役试验方案优化

安全壳整体试验是压水堆核电机组一项特大型、高风险、高难度的试验,通过模拟设计基准事故工况下安全壳内的峰值压力,在事故峰值压力平台下,进行安全壳整体泄漏率测量及各压力平台安全壳结构试验,以验证其密封和结构性能。安全壳整体试验是国家核安全局监管的一个重要见证点,试验结果直接决定是否能够启动反应堆发电。301大修安全壳整体试验是3号机组首次在役试验,本次试验汲取了秦山第二核电厂以往6次安全壳整体试验的经验和其他电厂的反馈,试验方案更加科学,试验的组织管理更为规范。文章对301大修安全壳整体试验的经验进行了论述和总结,希望对电厂以后的安全壳整体试验提供参考。     

Leakage behavior of inflatable seals subject to severe accident conditions

The paper describes tests to determine the leakage behavior of inflatable seals when subjected to containment pressures that exceed the design basis.² Inflatable seals are used to prevent leakage around personnel and escape lock doors in about 10% of the commercial nuclear power plant containment structures in the United States. All of the installations are in either Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) Mark-Ill type containments. This work is a part of an overall effort at Sandia National Laboratories to develop proven techniques for evaluating the performance of Light Water Reactor (LWR) containment buildings for beyond design basis loadings.Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Parameters that were monitored and recorded during each test were the internal seal pressure and temperature, chamber (containment) pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. An empirically based, analytical method is presented to predict the containment pressure at which significant leakage past inflatable seals can be expected.     

The relevance of IAEA tests to severe accidents in nuclear fuel cycle transport

Abstract

The design and performance standards for packages used for the transport of nuclear fuel cycle materials are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials, TS-R-1, in order to ensure safety under both normal and accident conditions of transport. The underlying philosophy is that safety is vested principally in the package and the design and performance criteria are related to the potential hazard. Type B packages are high-duty packages which are used for the transport of the more radioactive materials, notably spent fuel and vitrified high-level waste (VHLW). Tests are specified in the IAEA regulations to ensure the integrity of these packages in potential transport accidents involving impacts, fires or immersion in water. The mechanical tests for Type B packages include drop tests onto an unyielding surface without giving rise to a significant release of radioactivity. The objects which could impact upon a package in real-life transport accidents, such as concrete roads, bridge abutments and piers, will yield to some extent and absorb some of the energy of the moving package. Impact tests onto an unyielding surface are therefore relevant to impacts onto real-life objects at much higher speeds. The thermal test specifies that Type B packages must be able to withstand a fully engulfing fire of 800°C for 30 min without significant release of radioactivity, and this has to be demonstrated, for example, by analytical studies backed up by experimental tests. The regulations also specify immersion tests for Type B packages of 15 m for 8 h without significant release of radioactivity; and in addition for spent fuel and VHLW packages, 200 m for 1 h without rupture of the containment. There is a large body of evidence to show that the current IAEA Type B test requirements are severe and cover all the situations which can be realistically envisaged in the transport of spent fuel, VHLW and other fuel cycle materials. Any proposals for more severe tests, which have little technical justification, should therefore be treated with caution since this could result in a loss of public confidence in the current regulations, and the ratcheting up of design requirements which could not be justified on quantitative safety grounds.     

Experiments to evaluate behavior of containment piping bellows under severe accident conditions

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, NM. Several different bellows geometries representative of actual containment bellows are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of 13 tests have been conducted. The tests showed that bellows are capable of withstanding relatively large deformations up to or near the point of full compression before developing leakage. The test data are presented and discussed.     

High-temperature leak-characteristics of PCV hatch flange gasket

Small-model tests were performed to examine the integrity of the containment flange gasket in a severe accident. During a severe accident, containment structures suffer slow pressurization at relatively high temperatures. A realistic understanding of containment performance in such conditions is a major concern in developing an accident management strategy. This paper describes the results of experiments on the sealing capability of flange gaskets at high pressures and high temperatures. Silicone-rubber gaskets, which are used as the sealing material in BWR plant primary containment vessels (PCV) in Japan, were examined in small-model tests. The gaskets show sufficient sealing capability up to 225°C at 20 kgf/cm². When applying the leakage characteristics specified in this paper to codes for severe accidents, the results should be examined carefully based on realistic heattransfer phenomena.     

Results of large-scale test of discontinuity region in a prestressed concrete containment

The test described in this paper is part of an Electric Power Research Institute (EPRI) program (Research Program RP2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Phase 2 of the EPRI program, on which this paper is based, includes tests of five large-scale specimens with steel liner plates. The specimens represent structural elements of prestressed concrete containment buildings. Four full-scale square wall element specimens and one specimen representing the wall/basemat junction region were tested. This paper describes results of the wall/basemat junction region test.Results of this experimental work indicate that highly localized strains in the steel liner plate caused by internal overpressurization or other accident conditions can result in liner tearing and subsequent containment leakage. It appears that this liner tearing occurs in a controller manner. Extrapolating from these test results, leakage and depressurization is more likely to occur than global failure.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

福清核电厂1、2号机组安全壳整体泄漏率试验充压和降压速率优化的分析和研究

福清核电厂1、2号机组安全壳整体泄漏率试验在换料大修期间进行并作为大修关键路径的主线工作,而安全壳整体泄漏率试验的充压和降压速率是影响主线工作的重要因素。为提高机组的可利用率和经济性,基于法规和标准的要求,结合同类型机组的经验反馈对安全壳整体泄漏率试验进行优化,分析和研究试验中充压和降压速率提升的可行性。     

Testing of large-scale concrete containment structural elements

The tests described in this paper are part of an Electric Power Research Institute (EPRI) program (Research Project 2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Experimental study in Phase 2 of the investigation, on which this paper is based, includes tests of five large-scale specimens with steel liner plates representing structural elements of prestressed concrete containment buildings. Four square wall element specimens and one specimen representing the wall/basemat junction region were tested.This experimental work indicates that under internal overpressurization or other accident conditions, highly localized strains in the steel liner plate can result in liner tearing and subsequent containment leakage. These results support the theory of leak before break where liner tearing occurs in a controlled manner and leakage and depressurization occur rather than global failure.