ACME整体性能试验设施工作压力选取方案分析

拟建造的先进堆芯冷却机理试验台架（ACME）是验证压水堆核电站非能动安全系统性能及其安全分析软件的整体性能试验设施。在介绍AP1000电站整体性能试验台架及其评价的基础上，分析了不同工作压力对试验的影响。重点阐述了ACME工作压力的选取方案及其特点，探讨了确定试验初始状态的方法。分析表明：选取9.3MPa作为ACME的工作压力，涵盖了主要非能动系统工作的压力范围，具有等压等物性模拟非能动压水堆电站LOCA等事故工况的能力，是一个先进的非能动堆芯冷却整体性能试验设施设计方案。     

Estimation of the common cause failure probabilities of the components under mixed testing schemes

For the case where trains or channels of standby safety systems consisting of more than two redundant components are tested in a staggered manner, the standby safety components within a train can be tested simultaneously or consecutively. In this case, mixed testing schemes, staggered and non-staggered testing schemes, are used for testing the components. Approximate formulas, based on the basic parameter method, were developed for the estimation of the common cause failure (CCF) probabilities of the components under mixed testing schemes. The developed formulas were applied to the four redundant check valves of the auxiliary feed water system as a demonstration study for their appropriateness. For a comparison, we estimated the CCF probabilities of the four redundant check valves for the mixed, staggered, and non-staggered testing schemes. The CCF probabilities of the four redundant check valves for the mixed testing schemes were estimated to be higher than those for the staggered testing scheme, and lower than those for the non-staggered testing scheme.     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

水池贮源型γ辐照装置辐射安全设施分析

本文叙述了水池贮源型γ辐照装置安全设施的概况。汇总了国内一些现有辐照装置的具体安全设施。对照国家相关标准,对辐照装置的安全设施按其功能分项进行了讨论,并分析了安全联锁中可能存在的逻辑错误。最后强调指出,辐照装置安全设施的设计、安装者,不能把辐射安全问题交给管理者;而辐射安全管理不能因为有了良好的安全设施而有丝毫的放松。     

Operational analysis of nuclear research facilities in 1999–2008

The actual data on the number, type, operating state, and use of nuclear research facilities are presented. The generalized operational indices of the facilities for 1999–2008 are given; they were obtained on the basis of an analysis of the information entering the sector center for the collection and analysis of safety information concerning nuclear research facilities. Information is presented on the research being conducted at the facilities and the intensity with which the research reactors are used. Attention is focused mainly on the safety of nuclear research facilities. The results of an analysis of disruptions of the operation of the facilities are examined in detail. It is shown that the operation of nuclear research facilities is safe from the nuclear and radiological standpoints.     

对美国燃料循环设施抵御自然灾害标准的分析与思考

自然灾害威胁核设施的安全,福岛核事故的教训表明,核设施在选址、设计、建造中要进一步考虑自然灾害引发的外部事件对设施安全的影响.文章在对美国能源部关于燃料循环设施抵御自然灾害标准分析理解的基础上,提出制订我国核燃料循环设施抵御自然灾害标准的建议,以保证这些设施的安全.     

震后核设施性能综合检查与评价

5.12汶川大地震后,为及时评价地震对中国核动力研究设计院所属核设施造成的影响,采用检查(射线探伤、超声探伤、渗透探伤及水下视频检查等)、试验(功能、性能试验)、分析(抗震分析、断裂分析)以及审查确认等多种方法和手段对中国核动力研究设计院所属核设施进行了综合检查与评价,这是国内首次对民用核设施进行地震后的综合评价.主要的检查、评价结果及结论为:在检查范围内未发现汶川大地震对中国核动力研究设计院所属研究堆和临界装置造成损害,各研究堆和临界装置的安全停堆、冷却及限制放射性释放的三大基本安全功能仍得到保证.同时,建议继续开展厂址地震动研究工作,确定厂址地震设计基准;对应急计划进行修订,增加专项地震应急预案等.     

Refurbishment and safety upgradation of research reactor Cirus

Cirus, a 40 MW t, vertical tank type research reactor, having wide range of research facilities, was commissioned in the year 1960. This research reactor, situated at Mumbai, India has been operated and utilized extensively for isotope production, material testing and neutron beam research for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out during the early 1990s. Based on these studies, refurbishment of Cirus for its life extension was taken up. During refurbishment, additional safety features were incorporated in various systems to qualify them for the current safety standards. This paper gives the details of the operating experiences, utilization of the reactor along with methodologies followed for carrying out detailed ageing studies, refurbishment and safety upgradation for its life extension.     

核燃料循环设施物项安全分级方法研究

简单合理的物项安全分级,不仅可以提高设施的安全性,而且还可以减少审评双方的分歧,降低营运单位和设计单位的工作量。在分析国内外核动力装置采用核安全功能进行物项安全分级和乏燃料后处理设施采用剂量准则开展物项安全分级的基础上,研究提出了采用放射性物质包容量开展核燃料循环设施的物项安全分级的方法,并采用“未缓解释放”的事故分析方法,将放化安全一级(250 mSv)和放化安全二级(5 mSv)对应的剂量准则转化为放射性物质包容量限值。     

Comparison of nuclear pulse reactor facilities with reactivity-initiated-accident testing capability

Transient testing of advanced nuclear fuels and other structural materials is pivotal for the research, development, and ultimate demonstration of nuclear energy. Transient testing capabilities exist on a global scale, but these various facilities have different operations and design characteristics. Furthermore, the irradiation experiment vehicles (IEV) used in these transient reactor facilities have varying designs depending on the materials and experiment requirements. The advantages and disadvantages of each facility are presented and discussed by comparing the design and capabilities across nuclear transient reactor facilities (TRF). Further, a discussion of a specific IEV from each TRF shows the operational similarities and differences across TRFs. This comparison shows how some TRF designs are best suited for specific research areas and how others are being re-designed with flexible capabilities in mind. This inventory and comparison of global nuclear transient testing capabilities provides insight into the history, current status and future of nuclear fuel and technology development.     

Seismic re-evaluation of nuclear facilities worldwide: overview and status

Existing nuclear facilities throughout the world are being subjected to severe scrutiny of their safety in the event of an earthquake. In the United States, there have been several licensing and safety review issues for which industry and regulatory agencies have cooperated to develop rational and economically feasible criteria for resolving the issues. Currently, all operating nuclear power plants in the United States are conducting an Individual Plant Examination of External Events, including earthquakes beyond the design basis. About two-thirds of the operating plants are conducting parallel programs for verifying the seismic adequacy of equipment for the design basis earthquake. The U.S. Department of Energy is also beginning to perform detailed evaluations of their facilities, many of which had little or no seismic design. Western European countries also have been re-evaluating their older nuclear power plants for seismic events often adapting the criteria developed in the United States. With the change in the political systems in Eastern Europe, there is a strong emphasis from their Western European neighbors to evaluate and upgrade the safety of their operating nuclear power plants. Finally, nuclear facilities in Asia are also being evaluated for seismic vulnerabilities. This paper focuses on the methodologies that have been developed for re-evaluation of existing nuclear power plants and presents examples of the application of these methodologies to nuclear facilities worldwide.     

浮动核电站抑压水池液舱晃荡研究

浮动核电站运行环境与陆地核电站有着显著不同,其专设安全设施的设计需考虑海洋环境适应性,尤其是涉及液体流动的设施更应考察船体运动激励的影响。本文以抑压水池为分析对象,采用有限体积法对液舱晃荡过程进行仿真分析,研究在极限海洋环境下水池内水位的变化,以及有/无制流板情况下水位的区别。研究表明,船体纵/横摇引起的抑压水池最低水位相比于初始水位显著降低,由于抑压水池共振周期与船体晃荡周期错开,因此水位变化主要由晃荡幅值和内部结构件共同作用决定。      

Transportable nuclear power facilities in the INPRO International Project

Questions concerning safety, nonproliferation, monitoring of nuclear materials, civilian responsibility for nuclear risks, physical protection, transport operations, and others are analyzed within the framework of the INPRO project in application to transportable nuclear energy facilities. Essentially, the operative nuclear law and the experience of world nuclear power make it possible to solve the problems of the legal basis for the life cycle of transportable nuclear power facilities. To attain a system with the optimal accessibility, effectiveness, and safety, the nuclear power facilities will have to be adapted to the new specific requirements and conditions, and the international legal basis will have to be made more precise.     

SCC testing of pipe materials in BWR environment

A test loop has been installed in Ringhals 1 BWR, including facilities for Constant Elongation Rate Testing (CERT) and Electrochemical Potential (ECP) measurements in primary reactor water at reactor operation temperature. The loop is designed as to minimize transport time for reactor water from the reactor pressure vessel to the specimens being tested. Thus the testing environment is representative of the primary piping systems of BWRs, also with regard to short-lived constituents like hydrogen peroxide.The test program, which is in progress, has covered seven tests during start-up conditions or during power operation with presently current reactor water chemistry. In this presentation only CERT testing results on heavily sensitized austenitic chromium—nickel stainless steel are presented, although many other materials have been tested.Results show sensitized austenitic stainless steel is more prone to intergranular stress corrosion cracking (IGSCC) in actual than in simulated BWR environment and that start-up environment is chemically more aggressive than power operation environment. Reproducibility of the CERT technique as used is excellent.     

Test Facilities for Radioactive Material Transport Packages (AEA Technology,Winfrith, UK)

Abstract

Transport packages for radioactive materials are tested to demonstrate compliance with national and international regulations. The involvement of AEA Technology is traced from the establishment of the early IAEA Regulations. Transport package design, testing, assessment and approval requires a wide variety of skills and facilities. The comprehensive capability of AEA Technology in these areas is described with references to practical experience in the form of a short bibliography. The facilities described include drop-test cranes and targets (up to 700te); air guns for impacts up to sonic velocities; pool fires, furnaces and rigs for thermal tests including heat dissipation on prototype flasks; shielding facilities and instruments; criticality simulations and leak test instruments. These are illustrated with photographs demonstrating the comprehensive nature of package testing services supplied to customers.     

Consequence analysis for nuclear reactors,Yongbyon

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea.     

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt–water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb–Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed.     

The stage reached in the drafting of KTA-rules for the nondestructive testing of austenitic materials

At the end of 1972, the KTA (the Nuclear Commission) was called into being by the BMI (the Federal Ministry of the Interior) and by industry to draw up mandatory technical rules on dimensioning, manufacturing, design, and operating of nuclear facilities. The KTA-rules also lay down - where necessary - details of nondestructive testing. The procedure to be followed in the nondestructive testing of austenitic materials or weldings is the subject of ongoing discussion. In the following, consideration is given to the provisions of the KTA-rule 3201 currently in force and of proposed modifications or supplements concerning the NDT of austenitic materials or weldings on components in the coolant system for light water reactors. Final provisions especially for the ultrasonic examination of austenitic weld seams are still in preparation.     

放射性废物处置安全若干进展

结合2005年放射性废物处置安全国际大会反映出来的放射性废物处置安全领域的最新进展,介绍了全球放射性废物安全框架、废物处置安全战略、安全方案、地质处置设施安全、近地表处置设施安全、中等深度废物处置方案和公众沟通等方面的若干新进展和新观点。