Regulatory
Document
RD–310
Safety Analysis for Nuclear Power Plants February 2008
CNSC REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) develops regulatory documents under the authority of paragraphs 9(b) and 21(1)(e) of the Nuclear Safety and Control Act (NSCA).
Regulatory documents provide clarifications and additional details to the requirements set out in the NSCA and the regulations made under the NSCA, and are an integral part of the regulatory framework for nuclear activities in Canada.
Each regulatory document aims at disseminating objective regulatory information to stakeholders, including licensees, applicants, public interest groups and the public on a particular topic to promote consistency in the interpretation and implementation of regulatory requirements.
A CNSC regulatory document, or any part thereof, becomes a legal requirement when it is referenced in a licence or any other legally enforceable instrument.
Regulatory Document
RD–310
SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS
Published by the
Canadian Nuclear Safety Commission
February 2008
REGULATORY DOCUMENT
RD–310
SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS Published by the Canadian Nuclear Safety Commission
? Minister of Public Works and Government Services Canada 2008
Extracts from this document may be reproduced for individual use without permission, provided the source is fully acknowledged. However, reproduction in whole or in part for purposes of resale or redistribution requires prior written permission from the Canadian Nuclear Safety Commission.
Catalogue number: CC173-3/4-310E-PDF
ISBN 978-0-662-47989-5
Ce document est également disponible en fran?ais sous le titre Analyses de la s?reté pour les centrales nucléaires.
Document availability
The document can be viewed on the CNSC web site at www.nuclearsafety.gc.ca. Copies may be ordered in English or French using the contact information below:
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario, CANADA, K1P 5S9
Telephone: 613-995-5894 or 1-800-668-5284 (Canada only)
Facsimile: 613-992-2915
E-mail: info@cnsc-ccsn.gc.ca
PREFACE
This regulatory document was developed pursuant to the requirements and obligations set forth in the General Nuclear Safety and Control Regulations and in the Class I Nuclear Facilities Regulations, where a safety analysis report demonstrating the safety of the nuclear facility must be submitted to the Canadian Nuclear Safety Commission (CNSC).
The document identifies high-level regulatory information for a nuclear power plant licence applicant’s preparation and presentation of a safety analysis. The information required adheres to high standards and is consistent with modern national and international practices addressing issues and elements that control and enhance nuclear safety. In particular, it establishes a more modern risk-informed approach to the categorization of accidents, one that considers a full spectrum of possible events including the events of greatest consequence to the public.
The CNSC expects proponents and applicants for new reactor licences to immediately apply this regulatory document in new-build submissions. In the context of existing reactors, CNSC expects the licensees to apply this document, in a graduated manner, to all relevant programs in future submissions.
TABLE OF CONTENTS
1.0PURPOSE (1)
2.0SCOPE (1)
3.0RELEVANT LEGISLATION (1)
4.0SAFETY ANALYSIS OBJECTIVES (2)
5.0SAFETY ANALYSIS REQUIREMENTS (3)
5.1Responsibility (3)
5.2Events to be Analyzed (3)
5.2.1Identifying Events (3)
5.2.2Scope of Events (3)
5.2.3Classification of Events (4)
5.3Acceptance Criteria (4)
5.3.1Normal Operation (4)
5.3.2Anticipated Operational Occurrences and Design Basis Accidents (4)
5.3.3Beyond Design Basis Accidents (5)
5.3.4Acceptance Criteria for AOOs and DBAs (5)
5.4Safety Analysis Methods and Assumptions (5)
5.4.1General (5)
5.4.2Analysis Method (6)
5.4.3Analysis Data (6)
5.4.4Analysis Assumptions (7)
5.4.5Computer Codes (7)
5.4.6Conservatism in Analysis (7)
5.5Safety Analysis Documentation (7)
5.6Safety Analysis Review and Update (8)
5.6.1Review of Safety Analysis Results (8)
5.6.2Update of Safety Analysis (8)
5.7Quality of Safety Analysis (9)
GLOSSARY (11)
ASSOCIATED DOCUMENTS (13)
SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS
1.0 PURPOSE
The purpose of this regulatory document is to help assure that during the construction,
operation or decommissioning of a nuclear power plant (NPP), adequate safety analyses are completed by, or on behalf of, the applicant or licensee in accordance with the
Nuclear Safety and Control Act (NSCA) and regulatory requirements.
2.0 SCOPE
This document sets out the requirements related to safety analysis, including the selection of events to be analyzed, acceptance criteria, safety analysis methods, and safety analysis documentation and review.
LEGISLATION
3.0 RELEVANT
The relevance of the NSCA and the regulations made under the NSCA to this regulatory document is as follows:
1.Subsection 24(4) of the NSCA provides that the Commission may only issue, renew
or amend licences if the licensee or the applicant is (a) qualified to carry on the
activity that the licence authorizes the licensee to carry on, and (b), in carrying out
that activity, make adequate provision for the protection of the environment, the
health and safety of persons and the maintenance of national security and measures
required to implement international obligations to which Canada has agreed;
2.Subsection 24(5) of the NSCA authorizes the Commission to include in a licence
any term or condition that the Commission considers necessary for the purposes of
the Act;
3.Paragraph 3(1)(i) of the General Nuclear Safety and Control Regulations provides
that an application for a licence shall contain, in addition to other information, “a
description and the results of any test, analysis or calculation performed to
substantiate the information included in the application”;
4.Paragraph 5(f) of the Class I Nuclear Facilities Regulations provides that an
application for a licence to construct a Class I nuclear facility shall contain, in
addition to other information, information on “a preliminary safety analysis report
demonstrating the adequacy of the design of the nuclear facility”;
5.Paragraph 5(i) of the Class I Nuclear Facilities Regulations provides that an
application for a licence to construct a Class I nuclear facility shall contain, in
addition to other information, information on “the effects on the environment and
the health and safety of persons that may result from the construction, operation and
decommissioning of the nuclear facility…”;
6.Paragraph 6(c) of the Class I Nuclear Facilities Regulations provides that an
application for a licence to operate a Class I nuclear facility shall contain, in
addition to other information, information on “a final safety analysis report
demonstrating the adequacy of the design of the nuclear facility”;
7.Paragraph 6(h) of the Class I Nuclear Facilities Regulations provides that an
application for a licence to operate a Class I nuclear facility shall contain, in
addition to other information, information on “the effects on the environment and
the health and safety of persons that may result from the operation and
decommissioning of the nuclear facility…”; and
8.Paragraph 7(f) of the Class I Nuclear Facilities Regulations provides that an
application for a licence to decommission a Class I nuclear facility shall contain, in
addition to other information, information on “the effects on the environment and
the health and safety of persons that may result from the decommissioning of the
nuclear facility…”
OBJECTIVES
ANALYSIS
4.0 SAFETY
Safety analysis is an essential element of a safety assessment. It is an analytical study
used to demonstrate how safety requirements are met for a broad range of operating
conditions and various initiating events. Safety analysis involves deterministic and
probabilistic analyses in support of the siting, design, commissioning, operation or
decommissioning of a nuclear power plant. The requirements for probabilistic safety
assessment (PSA) for NPPs are provided in regulatory standard S-294, Probabilistic
Safety Assessment (PSA) for Nuclear Power Plants. Regulatory Document RD-310
focuses on the deterministic safety analysis used in the evaluation of event consequences.
The objectives of deterministic analysis are to:
1.Confirm that the design of a nuclear power plant meets design and safety
requirements;
2.Derive or confirm operational limits and conditions that are consistent with the
design and safety requirements for the NPP;
3.Assist in establishing and validating accident management procedures and
guidelines; and
4.Assist in demonstrating that safety goals, which may be established to limit the risks
posed by the nuclear power plant, are met.
This document identifies high-level requirements for conducting and presenting a safety analysis, taking into account best national and international practices.
5.0 SAFETY ANALYSIS REQUIREMENTS
5.1 Responsibility
The licensee is responsible for ensuring that the safety analysis meets all regulatory
requirements. The licensee shall:
1.
Maintain adequate capability to perform or procure safety analysis; 2.
Establish a formal process to assess and update safety analysis, which takes into account operational experience, research findings and identified safety issues; and 3. Establish and apply a formal quality assurance (QA) process that meets the QA
standards established for safety analysis in Canadian Standards Association (CSA)
publication N286.7-99, Quality Assurance of Analytical, Scientific and Design
Computer Programs for Nuclear Power Plants .
5.2 Events to be Analyzed
5.2.1 Identifying Events
The licensee shall use a systematic process to identify events, event sequences, and event combinations (“events” hereafter in this document) that can potentially challenge the
safety or control functions of the NPP. This process shall be based on regulatory
requirements and guidance, past licensing precedents, operational experience,
engineering judgment, results of deterministic and probabilistic assessments, and any
other systematic reviews of the design.
The identification of events shall account for all operating modes, and the list of
identified events shall be reviewed for completeness during the design and analysis
process and modified as necessary.
In addition to events that could challenge the safety or control functions of the NPP,
safety analysis shall be performed for normal operation.
5.2.2 Scope of Events
The list of events identified for the safety analysis shall include all credible:
1.
Component and system failures or malfunctions; 2.
Operator errors; and 3. Common-cause internally and externally initiated events.
A cut-off frequency shall be selected so that events with a frequency of occurrence less
than the cut-off limit provide only a negligible contribution to the overall risk posed by
the NPP. The elimination of such events from the analysis scope shall be justified and the reasons for eliminating them documented.
5.2.3 Classification of Events
The identified events shall be classified, based on the results of probabilistic studies and engineering judgement, into the following three classes of events:
1.Anticipated Operational Occurrences (AOOs) include all events with frequencies of
occurrence equal to or greater than 10-2 per reactor year;
2.Design Basis Accidents (DBAs) include events with frequencies of occurrence
equal to or greater than 10-5 per reactor year but less than 10-2 per reactor year; and
3.Beyond Design Basis Accidents (BDBAs) include events with frequencies of
occurrence less than 10-5 per reactor year.
Other factors to be considered in the event classification are any relevant regulatory
requirements or historical practices. Events with a frequency on the border between two classes of events, or with substantial uncertainty over the predicted event frequency, shall be classified into the higher frequency class.
Credible common-cause events shall also be classified within the AOO, DBA and BDBA classes.
Criteria
5.3 Acceptance
Operation
5.3.1 Normal
Analysis for normal operation of the NPP, performed during the design phase, shall
demonstrate that:
1.Radiological doses to workers and members of the public are within the limits
acceptable to the CNSC; and
2.Releases of radioactive material into the environment fall within the allowable
limits for normal operation.
5.3.2 Anticipated Operational Occurrences and Design Basis Accidents
Analysis for AOOs and DBAs shall demonstrate that:
1.Radiological doses to members of the public do not exceed the established limits;
and
2.The derived acceptance criteria, established in accordance with 5.
3.4 are met.
Basis Accidents
Design
5.3.3 Beyond
Analysis for BDBAs shall be performed as part of the safety assessment to demonstrate that:
1.The nuclear power plant as designed can meet the established safety goals; and
2.The accident management program and design provisions, put in place to handle the
accident management needs, are effective.
5.3.4 Acceptance Criteria for AOOs and DBAs
Qualitative acceptance criteria shall be established for each AOO and DBA to confirm
the effectiveness of plant systems in maintaining the integrity of physical barriers against releases of radioactive material. These qualitative acceptance criteria shall satisfy the
following general principles:
1.Avoid the potential for consequential failures resulting from an initiating event;
2.Maintain the structures, systems and components in a configuration that permits the
effective removal of residual heat;
3.Prevent development of complex configurations or physical phenomena that cannot
be modeled with high confidence; and
4.Be consistent with the design requirements for plant systems, structures and
components.
To demonstrate that these qualitative acceptance criteria applicable to the analyzed AOO or DBA are met, quantitative derived acceptance criteria shall be identified prior to
performing the analysis. Such derived acceptance criteria shall be supported by
experimental data.
The results of safety analysis shall meet appropriate derived acceptance criteria with
margins sufficient to accommodate uncertainties associated with the analysis.
The analysis shall be performed for the event that poses the most challenges in
demonstrating the meeting of derived acceptance criteria (i.e., the limiting event in an
event category).
5.4 Safety Analysis Methods and Assumptions
5.4.1 General
The analysis shall provide the appropriate level of confidence in demonstrating
conformity with the acceptance criteria. To achieve the appropriate level of confidence, the safety analysis shall:
1.Be performed by qualified analysts in accordance with an approved QA process;
2.Apply a systematic analysis method;
https://www.doczj.com/doc/343284936.html,e verified data;
https://www.doczj.com/doc/343284936.html,e justified assumptions;
https://www.doczj.com/doc/343284936.html,e verified and validated models and computer codes;
6.Build in a degree of conservatism; and
7.Be subjected to a review process.
Method
5.4.2 Analysis
The analysis method shall include the following elements:
1.Identifying the scenarios to be analyzed as required to attain the analysis objectives;
2.Identifying the applicable acceptance criteria, safety requirements, and limits;
3.Identifying the important phenomena of the analyzed event;
4.Selecting the computational methods or computer codes, models, and correlations
that have been validated for the intended applications;
5.Defining boundary and initial conditions;
6.Conducting calculations, including sensitivity cases, to predict the event transient,
starting from the initial steady state up to the pre-defined end-state;
7.Accounting for uncertainties in the analysis data and models;
8.Verifying calculation results for physical and logical consistency; and
9.Processing and documenting the results of calculations to demonstrate conformance
with the acceptance criteria.
5.4.3 Analysis
Data
The safety analysis shall be based on complete and accurate design and operational information.
The boundary and initial conditions used as the analysis input data shall:
1.Reflect accurately the NPP configuration;
2.Account for the effects of aging of systems, structures and components;
3.Account for various permissible operating modes; and
4.Be supported by experimental data, where operational data is not available. Significant uncertainties in analysis data, including those associated with nuclear power plant performance, operational measurements, and modeling parameters, shall be identified.
5.4.4 Analysis
Assumptions
Assumptions made to simplify the analysis, as well as assumptions concerning the
operating mode of the nuclear power plant, the availability and performance of the
systems, and operator actions, shall be identified and justified.
The analysis of AOO and DBA shall:
1.Apply the single-failure criterion to all safety systems and their support systems;
2.Account for consequential failures that may occur as a result of the initiating event;
3.Credit actions of systems only when the systems are qualified for the accident
conditions, or when their actions could have a detrimental effect on the
consequences of the analyzed accident;
4.Account for the possibility of the equipment being taken out of service for
maintenance; and
5.Credit operator actions only when there are
a)unambiguous indications of the need for such actions,
b)adequate procedures and sufficient time to perform the required actions, and
c)environmental conditions that do not prohibit such actions.
For the analysis of BDBA, it is acceptable to use a more realistic analysis methodology consisting of assumptions which reflect the likely plant configuration, and the expected response of plant systems and operators in the analysed accident.
Codes
5.4.5 Computer
Computer codes used in the safety analysis shall be developed, validated, and used in
accordance with a quality assurance program that meets the requirements of CSA
N286.7-99.
5.4.6 Conservatism in Analysis
The safety analysis shall build in a degree of conservatism to off-set any uncertainties
associated with both NPP initial and boundary conditions and modeling of nuclear power plant performance in the analyzed event. This conservatism shall depend on event class, and shall be commensurate with the analysis objectives.
5.5 Safety Analysis Documentation
The safety analysis documentation shall be comprehensive and sufficiently detailed to
allow for a conclusive review. The document shall include:
1.The technical basis for the analyzed event and key phenomena and processes;
2. A description of the analyzed facility, including important systems and their
performance, as well as operator actions;
https://www.doczj.com/doc/343284936.html,rmation describing the analysis method and assumptions;
4. A description of the assessments of code applicability for the analyzed event and
computer code uncertainty; and
5. A description of the analysis results in a manner that facilitates their understanding
and the drawing of conclusions related to conformance with acceptance criteria.
Analysis documentation shall facilitate the update of the analysis when new results
become available.
5.6 Safety Analysis Review and Update
5.6.1 Review of Safety Analysis Results
The licensee shall systematically review the safety analysis results to ensure that they are correct and meet the objectives set for the analysis. The results shall be assessed against the relevant requirements, applicable experimental data, expert judgment, and
comparison with similar calculations and sensitivity analyses.
The licensee shall review the analysis results using one or more of the following
techniques, depending on the objectives of the analysis:
1.Supervisory review;
2.Peer review;
3.Independent review by qualified individuals; and
4.Independent calculations using alternate tools and methods to the extent practicable.
5.6.2 Update of Safety Analysis
The safety analysis shall be periodically reviewed and updated to account for changes in NPP configuration, conditions (including those due to aging), operating parameters and procedures, research findings, and advances in knowledge and understanding of physical phenomena, in accordance with CNSC regulatory standard S-99, Reporting Requirements for Operating Nuclear Power Plants.
In addition to periodic updates, the safety analysis shall also be updated following the
discovery of information that may reveal a hazard that is different in nature, greater in
probability, or greater in magnitude than was previously presented to the CNSC in the
licensing documents.
5.7 Quality of Safety Analysis
Safety analysis shall be subject to a comprehensive QA program applied to all activities affecting the quality of the results. The QA program shall identify the quality assurance standards to be applied and shall include documented procedures and instructions for the complete safety analysis process, including, but not limited to:
1.Collection and verification of NPP data;
2.Verification of the computer input data;
3.Validation of NPP and analytical models;
4.Assessment of simulation results; and
5.Documentation of analysis results.
GLOSSARY
Acceptance criteria
Specified bounds on the value of a functional or conditional indicator used to assess the ability of a system, structure or component to meet its design and safety requirements.
Accident
Any unintended event, including operating errors, equipment failures or other mishaps, the consequences or potential consequences of which are not negligible from the point of view of protection or safety.
Anticipated operational occurrence (AOO)
An operational process deviating from normal operation that is expected to occur once or several times during the operating lifetime of the NPP but which, in view of the appropriate design provisions, does not cause any significant damage to items important to safety nor lead to accident conditions.
Best estimate method
A method designed to give realistic results.
Beyond design basis accident (BDBA)
Accident conditions less frequent and more severe than a design basis accident. A BDBA may or may not involve core degradation.
Common cause
A cause for a concurrent failure of two or more structures, systems or components, such as natural phenomena (earthquakes, tornadoes, floods, etc.), design deficiency, manufacturing flaws, operation and maintenance errors, human-induced destructive events and others. Conservative method
A method deliberately leading to results that are intended to be limiting relative to specified acceptance criteria.
Design basis accident (DBA)
Accident conditions against which an NPP is designed according to established design criteria, and for which the damage to the fuel and the release of radioactive material are kept within authorized limits.
Deterministic safety analysis
An analysis of nuclear power plant responses to an event, performed using predetermined rules and assumptions (e.g., those concerning the initial operational state, availability and performance of the systems and operator actions). Deterministic analysis can use either conservative or best estimate methods.
Event category
A group of events characterized by the same, or similar, cause and similarity in the governing phenomena.
Normal operation
Operation of an NPP within specified operational limits and conditions including start-up, power operation, shutting down, shutdown, maintenance, testing and refuelling.
NPP
See nuclear power plant.
Nuclear power plant
Also referred to as an NPP, a nuclear power plant is any fission-reactor installation that has been constructed to generate electricity on a commercial scale. A nuclear power plant is a Class IA nuclear facility, as defined in the Class I Nuclear Facilities Regulations.
Operational limits and conditions
A set of rules setting forth parameter limits or conditions that ensures the functional capability and the performance levels of equipment for safe operation of an NPP.
Operational mode
Operational mode may include start-up, operation at various power levels, shutting down, shutdown, maintenance, testing and refuelling.
Safety analysis
Evaluation of the potential hazards associated with the conduct of a proposed activity.
Safety assessment
Assessment of all aspects of the siting, design, commissioning, operation or decommissioning of an authorized facility that is relevant to safety.
Sensitivity analysis
A quantitative examination of how the behaviour of a system varies with change, usually in the values of the governing parameters.
Single-failure criterion
The criterion used to determine whether a system is capable of performing its function in the presence of a single failure.
Structures, systems and components (SSCs)
A general term encompassing all of the elements (items) of a facility or activity which contribute to protection and safety, except human factors.
Support features of safety systems
The collection of equipment that provides services such as cooling, lubrication and energy supply required by the protection system and the safety actuation systems.
February 2008 RD–310
ASSOCIATED DOCUMENTS
1.Probabilistic Safety Assessment (PSA) for Nuclear Power Plants, S-294. Canadian Nuclear
Safety Commission, Ottawa, 2005.
2.Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear
Power Plants. N286.7-99. Canadian Standards Association, 2003.
3.Reporting Requirements for Operating Nuclear Power Plants, S-99. Canadian Nuclear
Safety Commission, Ottawa, 2003.
13
核电站安全性分析姓名：X X X 学号：0 9 X X X X X X 专业：核工程与核技术 学院：核工程与地球物理学院 指导老师：X X
2012 年06月10 日 核电站安全性分析 东华理工大学核工系XXX 摘要：能源是社会和经济发展的基础，是人类生活和生产的要素。随着社会的发展，能源的需求也在不断扩大。从能源的供应结构来看，目前世界上消耗的能源主要来自煤、石油、天然气三大资源，这三种能源不仅利用率低，而且对生态环境造成严重污染。为了缓解能源矛盾，除了应积极开发太阳能、风能、潮汐能以及生物质能等再生资源外，核能是被公认的唯一实现的可大规模替代常规能源的即清洁又经济的现代能源。核能不仅单位能量大，而且资源丰富。地球蕴藏的铀矿和钍矿资源相当于有机燃料的几十倍。如果进一步实现控核聚变，并在海水中提取氚加以利用，就会从根本上解决能源供应矛盾。然而随着一系列的核事故的发生，核能的安全性再一步受到人们的质疑，本文简要回顾核电的发展，并对其安全性做了分析，指出核电是一种安全的能源。
关键词:能源核电安全 Nuclear power plant safety analysis East China University of Technology Nuclear Engineering XXX Abstract: Energy is the basis of the social and economic development, the elements of human life and production. With the social development, energy demand is also expanding. From the structure of energy supply, energy consumption in the world from the three resources of coal, oil, natural gas, three energy is not only a low utilization rate, and cause serious pollution to the ecological environment. In order to alleviate the energy contradictions, should actively develop solar, wind, tidal energy and biomass energy renewable resources, nuclear energy is recognized only can achieve large-scale alternative to conventional energy, clean and modern energy economy. Nuclear power units of energy, but also rich in natural resources. Global reserves of uranium and thorium mineral resources is equivalent to several times of the organic fuel. Further to achieve controlled nuclear fusion, and be used to extract tritium in seawater, will fundamentally solve the contradictions among the energy supply. However, with a series of nuclear accidents, the safety of nuclear energy and then step been questioned, briefly reviewed the development of nuclear power, and its
动火作业15种风险分析和安全措施。 1风险：易燃易爆有害物质 安全措施 ①将动火设备、管道内的物料清洗、置换，经分析合格。 ②储罐动火，清除易燃物，罐内盛满清水或惰性气体保护。 ③设备内通（氮气、水蒸气）保护。 ④塔内动火，将石棉布浸湿，铺在相邻两层塔盘上进行隔离。 ⑤进入受限空间动火，必须办理《受限空间作业证》。 2风险：火星窜入其他设备或易燃物侵入动火设备 安全措施 切断与动火设备相连通的设备管道并加盲板块隔断，挂牌，并办理《抽堵盲板作业证》。 3风险：动火点周围有易燃物 安全措施 ①清除动火点周围易燃物，动火附近的下水井、地漏、地沟、电缆沟等清除易燃后予封闭。
②电缆沟动火，清除沟内易燃气体、液体，必要时将沟两端隔绝。4风险：泄漏电流（感应电）危害 安全措施 5电焊回路线应搭接在焊件上，不得与其它设备搭接，禁止穿越下水道（井）。 6风险：火星飞溅 安全措施 ①高处动火办理《高处作业证》，并采取措施，防止火花飞溅。 ②注意火星飞溅方向，用水冲淋火星落点。 7风险：气瓶间距不足或放置不当 安全措施 ①氧气瓶、溶解乙炔气瓶间距不小于5m，二者与动火地点之间均不小于10m。 ②气瓶不准在烈日下曝晒，溶解乙炔气瓶禁止卧放。 8风险：电、气焊工具有缺陷 安全措施
动火作业前，应检查电、气焊工具，保证安全可靠，不准带病使用。9风险：作业过程中，易燃物外泄 安全措施 动火过程中，遇有跑料、串料和易燃气体，应立即停止动火。 10风险：通风不良 安全措施 ①室内动火，应将门窗打开，周围设备应遮盖，密封下水漏斗，清除油污，附近不得有用溶剂等易燃物质的清洗作业。 ②采用局部强制通风。 11风险：未定时监测 安全措施 ①取样与动火间隔不得超过30min，如超过此间隔或动火作业中断时间超过30min，必须重新取样分析。 ②做采样点应有代表性，特殊动火的分析样品应保留至动火结束。 ③动火过程中，中断动火时，现场不得留有余火，重新动火前应认真检查现场条件是否有变化，如有变化，不得动火。
核安全法规 HAF003（91） 核电厂质量保证安全法规 （1991年7月27日国家核安全局令第1号发布 1991年修改） 本规定是中华人民国核电厂安全法规的第四部分 本规定自一九九一年七月二十七日起实施 本规定由国家核安全局负责解释
目录 第一章引言 (3) 1．1 概述 (3) 1．2 围 (3) 1．3 责任 (4) 第二章质量保证大纲 (4) 2．1 概述 (4) 2．2 程序、细则及图纸 (5) 2．3 管理部门审查 (5) 第三章组织 (6) 3．1责任、权限和联络 (6) 3．2 单位间的工作接口 (6) 3．3 人员配备与培训 (7) 第四章文件控制 (7) 4．1 文件的编制、审核和批准 (7) 4．2 文件的发布和分布 (7) 4．3文件变更的控制 (7) 第五章设计控制 (8) 5．1 概述 (8) 5．2 设计接口的控制 (8) 5．3 设计控制 (8) 5．4 设计的变更 (8) 第六章采购控制 (9) 6．1 概述 (9) 6．2 对供方的评价和选择 (10) 6．3 对所购物项和服务的控制 (10) 第七章物项控制 (10) 7．1 材料、零件和部件的标识和控制 (10) 7．2 装卸、贮存和运输 (11) 7．3维护 (11) 第八章工艺过程 (11) 第九章检查和试验控制 (11) 9．1 检查大纲 (11) 9．2 试验大纲 (12) 9．3 测量和试验设备的标定和控制 (12) 9．4 检查、试验和运行状态的显示 (13) 第十章对不符合项的控制 (13) 10．1概述 (13) 10．2 对不符合项的审查和处理 (13) 第十一章纠正措施 (14) 第十二章记录 (14) 12．1质量保证记录的编写 (14) 12．2 质量保证记录的收集、贮存和保管 (14) 第十三章监查 (15) 13．1 概述 (15) 13．2 监查的计划安排② (15)
核电厂安全壳隔离阀密封性检测与分析 文章结合某核电厂安全壳隔离阀密封性试验，介绍了直接测量法的试验原理、操作方法和验收标准，通过贯穿件隔离阀密封性检测实例对试验的实际操作过程进行了阐述，并对试验结果进行了具体的分析和研究 标签：安全壳；隔离阀；密封性试验 1 概论 某核电厂反应堆安全壳是一座由钢筋混凝土底板、立式预应力混凝土筒壁和准球型预应力混凝土穹顶三部分组成的封闭预应力混凝土结构。反应堆安全壳是为防止核反应堆在运行或发生事故时放射性物质外逸的密闭容器，也称反应堆保护外壳。核电站反应堆发生事故时会大量释放放射性物质，安全壳作为最后一道核安全屏障，能防止放射性物质扩散污染周围环境。同时，也常兼作反应堆厂房的围护结构，保护反应堆设备系统免受外界的不利影响，它是一种体态庞大的特种容器。 安全壳内外系统管道的连通是通过机械贯穿件来实现的，贯穿件套筒按要求有不同的直径和厚度，以适应所连接的设备及由它们所传递的机械载荷，贯穿件的套筒焊在一块较厚的环形板上，该环形板则焊在安全壳的钢衬里上。机械贯穿件在安全壳内外两侧根据具体情况分别设置隔离阀门，以保证安全壳的密闭性。 安全壳密封性试验的目的是模拟一回路失水事故工况下，验证安全壳的整体密封性。安全壳密封性试验可以分为A类、B类、C类。其中A类试验是指安全壳整体的密封性试验，B类试验是指设备闸门、人员闸门、燃料通道和电气贯穿件的密封性试验，C类试验是指安全壳上所有的机械贯穿件试验，即贯件壳内外隔离阀的密封性试验。一般来说，C类局部密封性试验在每次换料冷停堆时进行，仅有个别十分可靠的机械贯穿件密封性试验每5个换料周期或10个换料周期进行一次。 2 试验原理和方法 直接测量法是安全壳隔离阀密封性试验的一种检测方法，流量补充法和流量收集法，都采用局部加压方式。试验对象为安全壳内外两侧隔离阀以及位于隔离阀和安全壳之间的支路阀门。试验时，气源通过局部检漏仪向隔离阀和边界阀门之间的管道内充压，由施加压力的方向与隔离阀在执行安全功能时受压方向相同，压力达到安全壳设计压力并稳定后，在保持压力不变的情况下，局部检漏仪通过向管道内补充的气体流量与压力的关系计算出隔离阀的泄漏率，并在显示屏中直接给出结果，单位为Nm3/h。流量收集法试验介质一般为水，试验方法相同，采用液态检漏仪加压，可以在试验阀门下游较低点通过容器收集方法确定该阀门的泄漏率。具体检测方法如下：
核电厂概率安全评价（PSA）技术研究参考文 本 In The Actual Work Production Management, In Order To Ensure The Smooth Progress Of The Process, And Consider The Relationship Between Each Link, The Specific Requirements Of Each Link To Achieve Risk Control And Planning 某某管理中心 XX年XX月
核电厂概率安全评价（PSA）技术研究 参考文本 使用指引：此安全管理资料应用在实际工作生产管理中为了保障过程顺利推进，同时考虑各个环节之间的关系，每个环节实现的具体要求而进行的风险控制与规划，并将危害降低到最小，文档经过下载可进行自定义修改，请根据实际需求进行调整与使用。 核电被称为技术设备、人的群体和组织三类元素的大 型经济实体，属科技密集型产业。对于核电厂而言，安全 是核电存在和发展的基础。在核电厂以往的系统安全分析 中，难以确定出具体的安全风险目标，在风险和费用之间 的权衡存在困难，更不易对事故发展的潜在原因及事故发 展的可能进程进行分析研究。基于此目的，概率安全评价 （PSA：Probability Safety Assessment）的提出，在系 统设计、制造、使用和维护的过程中，有力地支持了安全 风险的管理决策，保证了核电厂的安全运行。 1 PSA评价方法 1.1 概率论（PSA）方法
引入风险（risk）概论是为了比较和度量危险的大小和它们发生的可能性。PSA方法就是定量对核电厂作出其对环境造成各种风险的计算。PSA具有如下特点： 1）对所有事故谱（初因）进行评介； 2）对所有事故序列进行评价； 3）所有评价定量化。 核电厂PSA分成3个级别。一级，堆芯损坏分析：用事件树和故障树的概率方法，对设计和运行进行分析，得出导致堆芯熔化的事故序列及其发生频率；二级，源项分析：在一级分析的基础上分析事故的物理过程和安全壳的行为，计算不同事故释放类型的放射性源项；三级，后果评价；进行释出放射性物质特性、大气扩散程度和剂量评价。PSA评价的基本流程如图1所示。 1.2 初因的确定 首先要分析风险评价历史报告、反应堆运行历史的文
概率风险分析评价PRA又称为概率安全分析PSA，作为一种核安全评价方法，PSA 近年来发展很快。 作为一项评价技术，概率安全评价（PSA）用于找出复杂工程系统运行中所可能发生的潜在事故、估算其发生概率以及确定它们所可能导致的后果。概率安全评价是由安全性和统计学的概念在工程设计的应用中发展而来的。 概率安全评价(PSA)的应用可以追溯到上个世纪50年 代,最早应用于美国太空总署(NASA)的阿波罗登月计划, 1961年,美国贝尔实验室的H.A.Watson发展PSA的故障树 方法,将其应用于“民兵”导弹的发射控制系统的评估中,并 获得成功。1972年,PSA分析第1次应用于核电站设施上, 里程碑式的报告就是发表于1975年的W ASH-1400,分别用于 一个轻水堆和一个压水堆,开创了对于大型设备的安全进行 定量化描述的阶段。PSA用于工业辐照设备的安全分析开 始于90年代初[1-3],近年来取得较大发展。 1吴德强,译.国际放射防护委员会第76号出版物—潜在照射的 防护:对所选择辐射源的应用,北京:原子能出版社,1999. 2IAEA.Procedures for conductiong probabilistic safety assessment of nu- clear power plants(Level 1):A safety practice,safety series No.50-P-4, IAEA,Vienna.1992.
3IAEA.Human reliability analysis in probabilistic safety assessment for nuclear power plants,safety series No.50-P-10,IAEA,Vienna.1995. 安全评估分为动态和静态，以上可以放在最后 PRA,概率风险评价(PRA:ProbabilisticRisk Assessment) 自1972年美国原子能委员会(AEC)应用事件树和故障树相结合的分析技术成功地对核电站的风险进行了首次综合的评价,以定量 的方式给出了核电站的安全风险后,美国核管理委员会(NRC)开始使用PRA来支持其管理过程。在“挑战者”事件之后,NASA(美国航空航天局)制定了更严格的安全和质量保证大纲,采用概率评价方法对航天任务进行评价[2],并开发了一套完整的PRA程序对航天飞机的飞行任务进行评价, ESA(欧空局)的安全评价也从以定性为主转向定量评价,并开发了自己的风险评价程序[3]。PRA正作为许多工程系统安全风险管理程序的重要组成部分而应用于系统的设计、制造和使用运行中。 航天系统的安全性一直是人们所关注的问题。对航天系统进行安全性分析的方法经历了从定量到定性,再到定量的过程。早在50年代,美国
JSA风险分析、危害识别及安全措施河北瑞鹏建筑装饰工程有限公司
目录 一、范围与应用领域 二、参考文件 三、术语和定义 四、成立小组各成员职责 五、管理要求 六、程序流程 1、成立JSA小组 2、工作准备 3、危害辨识 4、风险评价
5、风险控制 一、范围与应用领域 1、目的 为规范工作安全分析，识别工作中每个工序、每个环节、每个阶段的风险因素，提出保护措施以消除风险或将风险降至可接受的程度，确保作业人员健康和安全，制定本程序。 2、适用范围 本程序适用于河北瑞鹏建筑装饰工程有限公司（以下简称“公司”)，所属的各施工项目部、临时施工现场、以及各施工班组。 二、参考文件 1、公司工作安全分析管理程序
2、GWDCD1/EMS205-2008 环境因素识别和评价控制程序 3、GWDCD1/HSEMS206-2008 对危害因素识别、风险评价和风险控制的策划控制程序 4、GWDCD1/HSEMS237-2008 健康安全与环境管理方案控制程序 5、GWDCD1/HSMS3023-2008 “两书一表”管理规定 三、术语和定义 1、工作安全分析（Job Safety Analysis简称JSA）注：以下正文中均使用该简称。 指：事先或定期对某项工作任务进行风险评价，并根据评价结果制定和实施相应的控制措施，达到最大限度消除或控制风险目的的方法。 2、暴露频率 每单位时间某事件发生的(或估计发生）次数。 3、严重性 可能引起的后果的严重程度。 4、可能性 后果事件发生的概率。 5、危害 能引起人员的伤害或对人员的健康（环境）造成负面影响的情况。（危害＝暴露频率×严重性） 6、风险 事件后果严重程度和发生的可能性的综合度量。（风险＝危害×可能性） 四、成立小组各成员职责 1、公司安全管理科组织制定、管理和维护本办法，其他职能部门和所属单位组织推行、实施本办法，并提供资源保障。
Regulatory Document RD–310 Safety Analysis for Nuclear Power Plants February 2008
CNSC REGULATORY DOCUMENTS The Canadian Nuclear Safety Commission (CNSC) develops regulatory documents under the authority of paragraphs 9(b) and 21(1)(e) of the Nuclear Safety and Control Act (NSCA). Regulatory documents provide clarifications and additional details to the requirements set out in the NSCA and the regulations made under the NSCA, and are an integral part of the regulatory framework for nuclear activities in Canada. Each regulatory document aims at disseminating objective regulatory information to stakeholders, including licensees, applicants, public interest groups and the public on a particular topic to promote consistency in the interpretation and implementation of regulatory requirements. A CNSC regulatory document, or any part thereof, becomes a legal requirement when it is referenced in a licence or any other legally enforceable instrument.
国家标准 《压水堆核电厂钢制安全壳设计建造规范》 编制说明 （征求意见稿） 标准编制组 2019年12月
压水堆核电厂钢制安全壳设计建造规范 一、任务来源及计划要求 本标准按照国家重点研发计划课题“基础通用与其它关键技术标准研究”（课题编号2017YFF0208004）任务书的要求以及与核工业标准化研究所签订的合同（合同编号为ISNI-KY-24-2019）内容进行编写。本标准由上海核工程研究设计院有限公司主编。 按照下达的计划，本标准计划于2019年12月31日前完成征求意见稿；于2020年3月31日前完成送审稿；于2020年6月30日前完成报批稿。 二、标准编制组组成 本标准由上海核工程研究设计院有限公司主编，编制组成员组成如下，详见表1。 表1：标准编制组成员名单 三、编制过程 3.1 总体过程 本标准的制定过程主要分为前期准备、征求意见稿编写阶段、送审稿编写阶段和报批稿编写阶段。 3.2 前期准备（2017年7月-2018年12月） 主要任务是成立标准编制小组，明确分工要求，分解工作任务、文件收集和调研分析、明确标准编制的进度控制。 在前期准备阶段成立标准编制小组和明确工作任务后，开展调研和文件收集工作。根据依托项目实施经验，确定了本标准编制的主要依据为ASME 规范NE 分卷，并参考国内压力容器设计规范（GB 150-2011）。此外还参考了相关的SRP 及RG导则进行规范的编制工作。
根据核电标准体系研究的前期工作分析结果，确定本标准的最初框架结构为：前言、目次、范围、术语和符号、总论、材料、设计、制造和安装、检测、试验、超压保护和附录。之后根据多次讨论和修改进行了必要的调整。 3.3 征求意见稿编写（2019年1月-2019年12月） 编制组在对参考文件进行详细分析的基础上，结合我国现状起草了本标准的工作组讨论稿，并在院内征求了专家意见。在具体章节编写过程中，对于标准内容的定位和合理安排问题征求了有关专家的意见，最终形成本标准征求意见稿。 3.4送审稿编写（2020年1月-2020年3月） 待广泛征求行业内的专家意见后，标准编写组将根据收到的专家意见对征求意见稿再进行深入地讨论，并对征求意见稿进行修改，按要求形成并提交送审稿。 3.5 报批稿编写（2020年4月-2020年6月） 届时根据标准《压水堆核电厂钢制安全壳设计建造规范》送审稿的审查情况，标准编写组将根据审查意见修改送审稿，完成了报批稿编写。 四、标准现状分析 我国监管机构国家核安全局批准出版的《核动力厂设计安全规定》（HAF 102-2016）和《核电厂反应堆安全壳系统的设计》（HAD102/06-1990），从法规和导则层面提出了核电厂安全壳系统和结构设计所需满足的要求，即要求在核电厂寿期内可能发生的所有荷载条件下，应保持安全壳结构的完整性和限制安全壳的泄漏。 因此，本标准在国内首次系统提出了钢制安全壳用材料技术要求、钢制安全壳设计技术要求和钢制安全壳建造技术要求。用于指导压水堆核电厂钢制安全壳设计、建造、试验和验收。 五、标准制修订背景和原则 5.1标准制修订背景 我国是从AP1000技术引进时，开始接触钢制安全壳这种设备，在之前的核电机组中，我国没有采用过钢制安全壳这种形式，国内相关法规及标准都未有涉及。当前我国第三代先进非能动核电站如依托项目（AP1000）、后续项目和示范项目（国和一号），都采用钢制安全壳设计。钢制安全壳是压水堆核电厂事故发生后的最后一道安全屏障，其功能包括余热排出及放射性废物的包容。在事故工
从福岛核电站事故分析看安全 文化(最新版) The core of safety culture is people-oriented, which requires the implementation of safety responsibilities in the specific work of all employees. ( 安全文化) 单位：_______________________ 部门：_______________________ 日期：_______________________ 本文档文字可以自由修改
从福岛核电站事故分析看安全文化(最新 版) 日本正遭遇二战以来最大的灾难，这次地震由于其史无前例的强烈震级和同时伴随的强次生灾害揪住了全球民众的心。这其中，福岛第一核电站事故1、2、3、4号机组所发生的事故，由于其可能对周边产生的恶劣影响和对人心理产生的恐慌，引起了越来越强烈的关注。根据诸多业内人士对核电站事故以及事故应急处理的分析，我们看到：福岛第一核电站事故看起来是天灾(地震引发海啸造成装置失效)，但其实也有许多人为因素，也就是说，还是有人做了不应该做的事情，有人没做应该做的事情。 下面我结合专业人士eagle506的技术分析谈一谈这其中的
文化因素。 1、关于应急处置 2011年3月11日下午，地震发生，反应堆安全停堆，按理应该马上向堆芯补水，保证堆芯冷却防止超压，但地震摧毁了电网，厂外电源不可用，这时应该发动应急柴油机，但海啸来了，柴油机房被淹，不过核电厂还备有蓄电池，虽然容量较小，但是在事故后8小时内还是为压力容器的冷却做了一些贡献的。电池眼看就要耗尽，为了保住压力容器，必须要卸压，防止压力容器超压爆炸。而且操作员也确实是这样做的。 但是，12日早，日本首相菅直人要来视察。 如果卸压，环境中的放射性会升高，虽然菅直人是空中视察，但这对没有穿防护服的日本首相来说仍然不是什么好事，所以，根据日本某些论坛的说法(没有得到官方证实)，卸压的事由于此次视察暂时中断。但余热不等人，安全壳内温度压力仍在上升。 菅直人走后，操作员开始继续释放压力容器内部的压力。此时压力容器内的温度约为550摄氏度，堆芯已经裸露并产生大
第3章核电厂事故分析的基本知识 3．1 核电厂事故分析的作用 事故分析是研究核电厂可能发生事故的种类及发生频率，确定事故发生后系统的响应及预计事故的进程，评价各种安全设施及安全屏障的有效性，研究各项因素及操纵员干预对事故进程的影响，估计事故情况下核电厂的放射性释放量及计算工作人员与居民所受的辐射剂量。 在核电厂设计过程中，事故分析用于选取停堆保护信号，确定停堆参数整定值和停堆延迟时间，确定缓解事故的专设安全设施的参数。 对于设计基准事件的分析是核电厂安全分析报告中必要的一章。分析的目的在于表明该核电厂设计足以控制这些事件的后果，使工作人员、公众和环境不至于受到不适当的放射性风险。 通过严重事故分析，可以找到核电厂的薄弱环节，有助于提高核电厂的安全性。严重事故分析，还可作为制定应急计划的依据。 3．2核电厂事故分析的方法 事故分析采用确定论及概率论方法，这两种方法相辅相成。设计基准事件的分析，以确定论方法为主；严重事故的分析，两种方法并用，侧重于概率论方法。 3．2．1确定论安全分析 从系统及部件失效和损坏，或人员失误的角度，假定事故确定地发生，按照分析问题的要求，选用保守或现实模型以及一系列规则和假设，分析计算整个核电厂系统的响应，直至得到该事故的放射性后果。 保守模型 又称评价模型。在分析中采用的初始条件及各项参数，均须从不利方面加上不确定性。要选用保守的各种关系式及标准，此外还必须考虑四项基本假设。保守模型一般用于核电厂安全审批过程，在该模型中考虑了最不利的情况，得出的是事故后果的极限值，给核电厂留有相当大的安全裕度。其缺点是分析所得的事故过程，有时与真实情况相差较远，使工作人员不能了解过程的实际变化。 现实模型 又称最佳估算模型。在分析中采用核电厂的运行参数或参数的平均值，尽量选用接近真实情况的关系式及标准，不考虑不合实际的保守假设。因而所得结果能接近真实情况。现实模型经常用于核电厂操作规程的制定和严重事故分析。作为一种尝试，目前正在研究使用现实模型分析，在其结果上加上适当裕度，作为代替保守模型或平行于保守模型的一种方法。 在用确定论方法进行事故分析中，所涉及的事故分析程序大致可分成以下六种。 (1)系统分析程序 可以模拟核电厂的一、二回路系统以及稳压器、蒸汽发生器、泵、阀门、燃料元件等设备。具有能计及各种反应性反馈的点堆或一维中子动力学模型，一般在流体力学上是一维的，有些程序堆芯是三维的，程序的规模大，一般有数万至20余万行。总体上分析核电厂在失水事故及各种瞬变过程中系统的响应，是事故分析中最主要的程序，如RETRAN，RELAP5，TRAC等。 (2)堆芯分析程序 或可称之为子通道分析程序，它以系统程序计算的结果作为边界条件，考虑堆芯内各处
浅谈核电站如何开展现场安全施工和检查安全是生产的首要任务，围绕安全开展的生产活动，才是我们作为设备检修工作的最终目的。所以，今天我们在这里讨论的是如何合理高效安全的开展核电站现场检修工作及如何做好现场工作的安全检查。 安全施工和安全检查工作是安全执行的重要步骤，是工作负责人识别现场不安全因素，按照程序或规程要求开展现场施工工作，而安全检查人员作为第三者去发现事故隐患并督促工作负责人改正存在的风险，事后做好安全教育和组织部门全员学习事件，做好此类事件的再次发生，但我们的目的就是做好风险预防工作，员工安全生产。所以下面从四个方面开展核电站的现场安全施工和安全检查工作： 其一、合理做好前期准备工作，落实安全风险管控； 每一项工作的开始都都少不了前期准备工作，而针对核电站开展现场生产活动，具备自己的特点。根据技术规格书来说，核电站的工作纷繁复杂，涉及有切割打磨、开工验电、容器（含地坑）内部作业、登高作业等风险作业，对应是火灾或爆炸风险、触电风险、人员窒息或中毒、高处坠落等事件发生都可以危机人身，严重时可导致人员死亡。 面对上面的提到或没有提到的风险，作为每项工作的准备人员，都要合理对每项工作进行全面细致的进行风险，做好每项工作的风险控制。在准备每项工作时，了解或查询类似工作，跟专业组负责人做好沟通交流，在条件的允许的前提下，安排人员提前了解现场情况。根据反馈情况，准备人员、专业组长及经验丰富的工作负责人，共同分析该项工作的风险分析，必要时与对口的科室班组的系统工程师和安全质量处的专项负责人进行沟通，制定合理有效的控制措施。在工作包中的风险分析填写完善的风险预估，编制安全施工组织设计或施工方案和风险分析，进一步办理工业安全高风险。编制同类事件的OE单，对工作负责人在开展工前会时，予以参考和借鉴。 总之，作为每项工作的发起者、组织者都要提前熟悉、了解现场工作存在的分析，提前做好风险分析，制定控制措施，为员工的安全生产做好铺垫。 其二、做好施工技术交底，严格执行施工要求； 每项工作的实施者是由我们在做的每位工作负责人执行，所以作为工作负责人，你们的责任重大。在开始执行的每一项工作，都要做好工前技术交底，开好每项工作的的工前会，对重大风险的工作的工前会，不但要通知部门管理责任人、准备工程师及电站级的安全员，还要通知质安部安全工程师、班组系统工程师或安全质量处的专业负责人，再次按照施工方案及OE单，梳理该项工作存在的风险，并对控制措施进行落实，做好人员分工，完善前期准备不足的地方。
编号：AQ-JS-08518 ( 安全技术) 单位：_____________________ 审批：_____________________ 日期：_____________________ WORD文档/ A4打印/ 可编辑 动火作业风险分析及安全对策 Risk analysis and safety countermeasures of hot work
动火作业风险分析及安全对策 使用备注：技术安全主要是通过对技术和安全本质性的再认识以提高对技术和安全的理解，进而形成更加科学的技术安全观，并在新技术安全观指引下改进安全技术和安全措施，最终达到提高安全性的目的。 序号 风险分析 安全对策 1 易燃易爆有害物质 1)将动火设备、管道内的物料清洗、置换，经分析合格。 2)储罐动火，清除易燃物，罐内盛满清水或惰性气体保护。 3)设备内通（氮气、水蒸气）保护。 4)塔内动火，将石棉布浸湿，铺在相邻两层塔盘上进行隔离。 5)进入受限空间动火，必须办理《受限空间作业证》。 2 火星窜入其它设备或易燃物侵入动火设备 切断与动火设备相连通的设备管道并加盲板可靠隔断，挂牌，
并办理《抽堵盲板作业证》。 3 动火点周围有易燃物 1)清除动火点周围易燃物，动火附近的下水井、地漏、地沟、电缆沟等清除易燃后予封闭。 2)电缆沟动火，清除沟内易燃气体、液体，必要时将沟两端隔绝。 4 泄漏电流（感应电）危害 电焊回路线应搭接在焊件上，不得与其它设备搭接，禁止穿越下水道（井）。 5 火星飞溅 1)高处动火办理《高处作业证》，并采取措施，防止火花飞溅。 2)注意火星飞溅方向，用水冲淋火星落点。 6 气瓶间距不足或放置不当
( 安全管理 ) 单位：_________________________ 姓名：_________________________ 日期：_________________________ 精品文档 / Word文档 / 文字可改 保证核电厂安全有哪些管理措 施(最新版) Safety management is an important part of production management. Safety and production are in the implementation process
保证核电厂安全有哪些管理措施(最新版) 管理措施之一——健全的国家监管机构 国家监管机构对核电厂实行全寿期监督管理，即从选址、设计、建造、调试、运行、直到退役和废物处理处置的各个环节。 我国民用核设施的核安全监督管理主要由国家核安全局负责。 管理措施之二——制定和完善核安全防护法规体系 国家有关部门发布实施核电厂厂址选择、设计、运行、质量保证、辐射防护和废物管理等安全规定以及辐射防护基本标准等，形成一整套比较完整的核安全、辐射防护法规标准体系。 管理措施之三——实行核设施安全许可证制度 核电厂在不同阶段，其营运单位要向国家核安全主管部门提交相应的报告。经审评，在条件完全符合国家有关规定后才颁发许可证。营运单位只有获得这些许可证后才能开展相应的工作。 管理措施之四——严密的质量保证体系
核电厂有严密的质量保证体系。对选址、设计、建造、调试、运行直至退役等各个阶段的每一项具体活动都有单项的质量保证大纲，并严格执行。 另外，还实行内部和外部监查制度，监督检查质量保证大纲的实施情况，确认起到应有的作用。例如，在建造阶段，要对设备进行监造，对施工进行监理。在运行阶段，要进行预防性检修、在役检查和定期试验，以保证机组的系统和设备的状态符合技术规范。 管理措施之五——对参与单位和人员严格要求 国家对参与核电厂建设的单位，甚至小到零部件制造单位，都要经审查合格后，方可开展相应的活动。 国家对参加核电厂工作的人员的选择、培训、考核和任命有严格的规定。以操纵员为例，要求选择基本素质好、有一定学历和工作经验的人员，经过课堂、核电厂模拟机和核电厂实际运行培训，再通过国家级的考试，领到操纵员执照后，才能上岗。上岗工作以后，还要定期考查和再培训，保证在工作岗位上的人员都合格。 管理措施之六——极其严密的安全保卫系统
核电厂的安全壳设计 1 引言 为了在电厂简化、安全性、可靠性、投资保护和电厂成本方面提供重大的、可度量的改进,AP1000采用非能动安全系统。安全壳是实现上述改进的一个关键设施。它不仅提供了防止裂变产物释放的高度完整、低泄漏率的屏障，其表面还承担将安全壳空气中的热量排到大气中去的传热功能。安全壳内部结构连同非能动堆芯冷却系统(PXS)和严重事故缓解设施一起设计。本文介绍了API000安全壳容器设计、建造、事故后特征和严重事故性能。此外，本文也讨论了放射性释放模式，假设条件以及安全壳短期和长期性能。 2 AP1000 安全壳设计概述 AP1000安全壳是一个自由直立的圆柱形钢制容器，带有椭球形的上封头和下封头。钢制安全壳容器被完全包容在一个混凝土屏蔽体中，该厂房提供了对外部危害(如飞射物)的防护，并限制中子、γ射线、散射照射对电厂工作人员和公众的辐照。 如图l 所示，API000设计保留了和AP600相同的直径，但其高度比AP600增加了7.8 m ，从而增加了自由空间。此外，与AP600相比，AP1000通过增加容器壁的厚度和使用SA738型B 级材料提高了安全壳的设计压力。AP1000安全壳容器的一些重要参数与AP600的比较见表l。如表中所示，圆柱形容器大部分的钢壁标称厚度是4.445cm，局部位置较厚，如设备闸门处。最低一层圈柱形壳体的壁厚增加到4.762 cm，以便为预埋件过渡段中的腐蚀情况留有裕度。封头是椭球形的，厚度为4.127cm，主直径为39.624m，而高度为11.468m。 安全壳容器由5个主要结构模块组装建造而成，每个模块都由预先成型的、喷好漆的钢板制成。为了进一步减少安全壳内的组装活动，这些模块包含环形加强筋、吊环梁、设备闸门、人员空气闸门、贯穿件组件和其它附件，其中包括非能动安全壳冷却系统(PCS)空气挡板的支撑和水分配溢流口的固定板。 安全壳容器的设计使其能支撑环吊及其载荷，并考虑了蒸汽发生器的更换。
核电厂概率安全评价（PSA）技术研究 核电被称为技术设备、人的群体和组织三类元素的大型经济实体，属科技密集型产业。对于核电厂而言，安全是核电存在和发展的基础。在核电厂以往的系统安全分析中，难以确定出具体的安全风险目标，在风险和费用之间的权衡存在困难，更不易对事故发展的潜在原因及事故发展的可能进程进行分析研究。基于此目的，概率安全评价（PSA：Probability Safety Assessment）的提出，在系统设计、制造、使用和维护的过程中，有力地支持了安全风险的管理决策，保证了核电厂的安全运行。 1 PSA评价方法 1.1 概率论（PSA）方法引入风险（risk）概论是为了比较和度量危险的大小和它们发生的可能性。PSA方法就是定量对核电厂作出其对环境造成各种风险的计算。PSA具有如下特点：1）对所有事故谱（初因）进行评介；2）对所有事故序列进行评价；3）所有评价定量化。核电厂PSA分成3个级别。一级，堆芯损坏分析：用事件树和故障树的概率方法，对设计和运行进行分析，得出导致堆芯熔化的事故序列及其发生频率；二级，源项分析：在一级分析的基础上分析事故的物理过程和安全壳的行为，计算不同事故释放类型的放射性源项；三级，后果评价；进行释出放射性物质特性、大气扩散程度和剂量评价。PSA评价的基本流程如图1所示。 1.2 初因的确定首先要分析风险评价历史报告、反应堆运行历史的文件资料以及作为PSA分析对象的核电厂设计资料进行工程判断，从中编制出初因事件的清单。在选择初因的过程中，要确定可能发生的事件，这些事件需要安全系统的投入以减缓后果并将反应堆带入安全状态。然后对事件进行分类，分类的准则是所需的系统响应和成功准则是否一致。图1 PSA 评价流程图初因事件的选择通常来源于以下几个方面：核电厂的个体情况；
编号：SM-ZD-44831 确保核电站安全的措施Through the process agreement to achieve a unified action policy for different people, so as to coordinate action, reduce blindness, and make the work orderly. 编制：____________________ 审核：____________________ 批准：____________________ 本文档下载后可任意修改
确保核电站安全的措施 简介：该方案资料适用于公司或组织通过合理化地制定计划，达成上下级或不同的人员之间形成统一的行动方针，明确执行目标，工作内容，执行方式，执行进度，从而使整体计划目标统一，行动协调，过程有条不紊。文档可直接下载或修改，使用时请详细阅读内容。 四道屏障 为防止放射性物质处逸设置了四道屏障： 1、燃料芯块； 2、密封的燃料包壳； 3、坚固的压力容器和密闭的一回路系统； 4、安全壳。 多重保护 在出现可能危及设备和人身安全的情况时： 1、进行正常停堆 2、因任何原因未能正常停堆时，控制棒自动落入堆内，实行自动紧急停堆； 3、如因任何原因控制棒未能插入，高深度硼酸水自动喷入堆内，实现自动紧急停堆。
对一切重要设备都采取了类似的多种保护措施，如设置了两路独立的可靠的外电源，当一路外电源因事故停电时，可自动切换到另一路供电。万一两路外电源同时断电怎么办？不要紧。核电站里还有由柴油发电机提供的紧急备用电源。 专设安全设施 人们常用“万无一失”来形容一件事物的安全可靠，而核电站为这极不可能出现的“一失”出作了周密准备，这就是专设安全设施。 我们可以设想这“一失”是管壁很厚的一回路主管道断裂了。这时专设安全设施投入工作，首先向堆内高压注水，防止堆内“烧干”；压力降低后，低村注水系统工作，继续向堆内注水冷却。与此同时，安全壳与外界自动隔离；安全部顶的喷淋系统自动喷淋冷水，降低安全壳的温度和压力；消氢系统投入工作，除去可能引起爆炸的氢气。 质量保证体系 核电站有着严密的质量保证体系。
AP1000的ATWS事故概率安全分析 概率安全分析能够对核反应堆事故发生过程进行全面分析,并可对潜在事故定量化。在核电厂安全分析中,作为确定论安全分析的补充可以识别出核电厂设计或运行的薄弱环节。 为了评价AP1000先进非能动型电厂在ATWS事故工况下的安全性能,本文对AP1000的ATWS事故进行了概率安全分析。主要研究内容和结论如下:论文将ATWS 分成三类:主给水不可用ATWS、安注信号已触发ATWS和主给水可用ATWS,分析了三类ATWS的事故进程和安全功能响应。 在此基础上,建立了三类ATWS的事件树。论文建立了缓解ATWS事故的 AP1000相关系统故障树,对故障树的不确定性和人因可靠性进行了详细分析,并深入地研究了共因失效。 给出了共因失效模型:Alpha因子模型、Beta因子模型和MGL模型在交错试验和非交错试验下的参数估计公式,评价了三种共因失效模型对系统失效概率的影响。论文应用Risk Spectrum软件完成了事故序列和系统故障树的定量和定性分析。 结果表明:AP1000的ATWS事故堆芯损坏频率均值为5.35E-10/ (堆·年),其90%置信度区间下限(5%)为3.70E-12/ (堆·年),上限(95%)为 1.55E-09/(堆·年)。研究给出了系统故障树和事故序列的最小割集及其发生概率;通过堆芯损坏重要度分析得出了风险和安全重要的基本事件;通过堆芯损坏敏感性分析,得到所有人因失误概率为1时,堆芯损坏频率为 2.52E-06/(堆·年),该数值比基准堆芯损坏频率增加了 4710倍,这说明缓解ATWS事故的人因行为是非常重要的;PRHR不可用导致堆芯损坏频率增加了 10.7倍,PRHR是非能动系统