27.120.10 反应堆工程 标准查询与下载

ISO 18229:2018 第四代核反应堆机械部件和金属结构的基本技术要求

This document defines the essential technical requirements that are addressed in the process of design and construction of Generation IV (GEN IV) nuclear reactors. It does not address operation, maintenance and in-service inspection of reactors. Six reactor concepts are considered for GEN IV: the sodium fast reactor, the lead fast reactor, the gas fast reactor, the very high temperature reactor, the supercritical water reactor and the molten salt reactor. Annex A details the main characteristics for the different concepts. The scope of application of this document is limited to mechanical components related to nuclear safety and to the prevention of the release of radioactive materials — that are considered to be important in terms of nuclear safety and operability, — that play a role in ensuring leaktightness, partitioning, guiding, securing and supporting, and — that contain and/or are in contact with fluids (such as vessels, pumps, valves, pipes, bellows, box structures, heat exchangers, handling and driving mechanisms).

Essential technical requirements for mechanical components and metallic structures foreseen for Generation IV nuclear reactors

ISO 19226:2017 核能. 反应堆容器和内部构件中子注量和每原子位移（DPA）的测定

This document provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity, between the ends of active fuel assemblies, given the neutron source in the core. NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production. The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs). This document also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

Nuclear energy - Determination of neutron fluence and displacement per atom (dpa) in reactor vessel and internals

ASTM C1725-17 热电池专用支持设备和工具的标准指南

1.1 Intent: 1.1.1 This guide presents practices and guidelines for the design and implementation of equipment and tools to assist assembly, disassembly, alignment, fastening, maintenance, or general handling of equipment in a hot cell. Operating in a remote hot cell environment significantly increases the difficulty and time required to perform a task compared to completing a similar task directly by hand. Successful specialized support equipment and tools minimize the required effort, reduce risks, and increase operating efficiencies. 1.2 Applicability: 1.2.1 This guide may apply to the design of specialized support equipment and tools anywhere it is remotely operated, maintained, and viewed through shielding windows or by other remote viewing systems. 1.2.2 Consideration should be given to the need for specialized support equipment and tools early in the design process. 1.2.3 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.3 Caveats: 1.3.1 This guide is generic in nature and addresses a wide range of remote working configurations. Other acceptable and proven international configurations exist and provide options for engineer and designer consideration. Specific designs are not a substitute for applied engineering skills, proven practices, or experience gained in any specific situation. 1.3.2 This guide does not supersede federal or state regulations, or both, or codes applicable to equipment under any conditions. 1.3.3 This guide does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. 1.4 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

Standard Guide for Hot Cell Specialized Support Equipment and Tools

NB/T 20470-2017 核电厂选址假想事故源项分析准则

Criteria for analysis of hypothetical accident source terms for nuclear power plant siting

NB/T 20143.3-2017 核空气和气体处理规范 工艺气体处理 第3部分： 放射性废气滞留设备

Nuclear Air and Gas Handling Code Process Gas Handling Part 3: Radioactive Waste Gas Retention Equipment

NB/T 20419-2017 压水堆核电厂安全壳过滤排放系统设计准则

本标准规定了压水堆核电厂安全壳过滤排放系统设计的基本要求，包括该系统的功能、系统设备、设计准则、试验和检查要求等。 本标准适用于压水堆核电厂安全壳过滤排放系统的设计。

Design criteria for containment filtration and exhaust system of pressurized water reactor nuclear power plant

DB34/T 2734.2-2016 托卡马克聚变堆部件 兼容性设计与评估 技术指南 第2部分：评估

DB34/T 2734 的本部分规定了聚变堆部件兼容性设计的评估方法和评估实施流程。 本部分适用于聚变堆部件设计的所有阶段。

Tokamak Fusion Reactor Component Compatibility Technical Guidelines for Design and Evaluation Part 2: Evaluation

DB34/T 2734.1-2016 托卡马克聚变堆部件 兼容性设计与评估 技术指南 第1部分：设计

DB34/T 2734 的本部分规定了托卡马克聚变堆部件关于维护需求与遥操作系统兼容性的设计要求。 本部分适用于托卡马克聚变堆部件设计的所有阶段.

Tokamak Fusion Reactor Component Compatibility Technical Guidelines for Design and Assessment Part 1: Design

KS C IEC 60557-2016 核反应堆领域的 IEC 术语

IEC terminology in the nuclear reactor field

KS C IEC 60557:2016 刷子架的定义和用语

IEC terminology in the nuclear reactor field

ASTM E2215-16 从轻水慢化核动力反应堆容器中评估监视胶囊的标准实施规程

1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules. 1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life. 1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel. 1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any lightwater moderated nuclear reactor.2 1.5 Modifications to the standard test program and supplemental tests are described in Guide E636. 1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

ASTM E185-16 轻水中等核反应堆船舶监控计划设计标准实践

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced lightwater small modular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2 ) at the inside surface of the ferritic steel reactor vessel. 1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. NOTE 1—The increased complexity of the requirements for a lightwater moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E185 describes the minimum requirements for design of a surveillance program. Practice E2215 describes the procedures for testing and evaluation of surveillance capsules removed from a reactor vessel. Guide E636 provides guidance for conducting additional mechanical tests. A summary of the many major revisions to Practice E185 since its original issuance is contained in Appendix X1. NOTE 2—This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. See Appendix X1.

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

ASTM E706-16 轻水反应堆压力容器监控标准标准主矩阵

1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel’s service life (Fig. 1). Referenced documents are listed in Section 2. The summary information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced standards (Section 2) and references for use by individual writers and users. More detailed writers’ and users’ information, justification, and specific requirements for the individual practices, guides, and methods are provided in Sections 3 – 5. General requirements of content and consistency are discussed in Section 6. 1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series. 1.3 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses (112)2 and Guide E509), neutron-induced changes in reactor pressure vessel steel fracture toughness must be predicted, then checked by extrapolation of surveillance program data during a vessel’s service life. Uncertainties in the predicting methodology can be significant. Techniques, variables, and uncertainties associated with the physical measurements of PV and support structure steel property changes are not considered in this master matrix, but elsewhere (2, 6, 7), (11-26), and Guide E509). 1.4 The techniques, variables and uncertainties related to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation procedures and data are addressed in separate standards belonging to this master matrix (1, 17). The main variables of concern to (1), (2), and (3) are as follows: 1.4.1 Steel chemical composition and microstructure, 1.4.2 Steel irradiation temperature, 1.4.3 Power plant configurations and dimensions, from the core periphery to surveillance positions and into the vessel and cavity walls. 1.4.4 Core power distribution, 1.4.5 Reactor operating history, 1.4.6 Reactor physics computations, 1.4.7 Selection of neutron exposure units, 1.4.8 Dosimetry measurements, 1.4.9 Neutron special effects, and 1.4.10 Neutron dose rate effects. 1.5 A number of methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads ((1, 7, 8, 11, 12, 14, 16, 17, 23-27), Referenced Documents: ASTM Standards (2.1), Nuclear Regulatory Documents (2.3) and ASME Standards (2.4)). As older LWR pressure vessels become more highly irradiated, the predictive capability for changes in toughness must improve. Since during a vessel’s service life an increasing amount of information will be available from test reactor and power reactor surveillance programs, procedures to evaluate and use this information must be used (1, 2, 4-9, 11, 12, 23-26, 28). This master matrix defines the current (1) scope, (2) areas of application, and (3) general grouping for the series of ASTM standards, as shown in Fig. 1. 1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.7 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the 1 This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05 on Nuclear Radiation Metrology. Current edition approved Dec. 1, 2016. Published January 2017. Originally approved in 1979. Last previous edition approved in 2002 as E0706 -2002 which was withdrawn July 2011 and reinstated in December 2016. DOI: 10.1520/E070616. 2 The boldface numbers in parentheses refer to a list of references at the end of this standard. Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee. 1 responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards

BS 4877:2016 核反应堆仪表与控制 实践守则

Nuclear reactor instrumentation and control. Code of practice

ASTM E798-16 在基于加速器的中子源进行辐照的标准实践

1.1 This practice covers procedures for irradiations at accelerator-based neutron sources. The discussion focuses on two types of sources, namely nearly monoenergetic 14-MeV neutrons from the deuterium-tritium T(d,n) interaction, and broad spectrum neutrons from stopping deuterium beams in thick beryllium or lithium targets. However, most of the recommendations also apply to other types of acceleratorbased sources, including spallation neutron sources (1).2 Interest in spallation sources has increased recently due to their development of high-power, high-flux sources for neutron scattering and their proposed use for transmutation of fission reactor waste (2). 1.2 Many of the experiments conducted using such neutron sources are intended to provide a simulation of irradiation in another neutron spectrum, for example, that from a DT fusion reaction. The word simulation is used here in a broad sense to imply an approximation of the relevant neutron irradiation environment. The degree of conformity can range from poor to nearly exact. In general, the intent of these experiments is to establish the fundamental relationships between irradiation or material parameters and the material response. The extrapolation of data from such experiments requires that the differences in neutron spectra be considered. 1.3 The procedures to be considered include methods for characterizing the accelerator beam and target, the irradiated sample, and the neutron flux (fluence rate) and spectrum, as well as procedures for recording and reporting irradiation data. 1.4 Other experimental problems, such as temperature control, are not included. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Practice for Conducting Irradiations at Accelerator-Based Neutron Sources

ASTM E521-16 使用带电粒子照射研究中子辐射损伤的影响的标准实践

1.1 This practice provides guidance on performing chargedparticle irradiations of metals and alloys, although many of the methods may also be applied to ceramic materials. It is generally confined to studies of microstructural and microchemical changes induced by ions of low-penetrating power that come to rest in the specimen. Density changes can be measured directly and changes in other properties can be inferred. This information can be used to estimate similar changes that would result from neutron irradiation. More generally, this information is of value in deducing the fundamental mechanisms of radiation damage for a wide range of materials and irradiation conditions. 1.2 Where it appears, the word “simulation” should be understood to imply an approximation of the relevant neutron irradiation environment for the purpose of elucidating damage mechanisms. The degree of conformity can range from poor to nearly exact. The intent is to produce a correspondence between one or more aspects of the neutron and charged particle irradiations such that fundamental relationships are established between irradiation or material parameters and the material response. 1.3 The practice appears as follows: Section Apparatus 4 Specimen Preparation 5 – 10 Irradiation Techniques (including Helium Injection) 11–12 Damage Calculations 13 Postirradiation Examination 14 – 16 Reporting of Results 17 Correlation and Interpretation 18 – 22 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Practice for Investigating the Effects of Neutron Radiation Damage Using Charged-Particle Irradiation

NB/T 20394-2016 核电厂安全级控制盘、屏和机架的设计与鉴定

本标准规定了核电厂安全级控制盘、屏和机架的设计和鉴定基本要求。 本标准适用于核电厂安全缀控制盘、屏和机架的设计与鉴定，其他核反应堆也可参照使用。 本标准不适用于安装在安全级控制盘、屏和机架上的单个部件、插件和外部敷设的电缆的设计与鉴定，但要考虑它们可能对安全级控制盘、屏和机架的设计与鉴定的影响。

Design and qualification of class 1E control boards, panels and racks in nuclear power plants

NB/T 20379-2016 核电厂安全相关的操纵员动作时间响应设计准则

本标准规定了轻水堆核电厂设计中核安全相关的操纵员动作时间响应设计准则，规定了安全分析中操纵员触发和控制核安全功能的可信动作应达到的时间要求。 本标准适用于轻水堆核电厂确定在缓解导致紧急停堆的设计基准事件中所涉及的核安全相关的操纵员动作的最小响应时间间隔。其它堆型可参考使用。

Time response design criteria for safety-related operator actions in nuclear power plants

NB/T 20393-2016 压水堆核电厂稳压器电加热器技术条件

本标准规定了压水堆核电厂稳压器电加热器的设计、制造、检验、试验、验收和包装运输等要求。 本标准适用于参照法国RCC系列规范建造的压水堆核电厂稳压器电加热器，其它压水堆核电厂稳压器电加热器可参照使用。

Specification of pressurizer electric heater for pressurized water reactor nuclear power plants

NB/T 25055-2016 核电厂汽轮机焊接转子检验规程

Inspection procedures for welded rotors of steam turbines in nuclear power plants