27.120.10 反应堆工程 标准查询与下载

GB 11850-1989 反应堆退役辐射防护规定

本标准规定了反应堆退役的辐射防护标准、原则、基本要求与措施。 本标准主要适用于生产堆的退役，也适用于研究试验堆的退役。

Regulation for radiation protection of reactor decommissioning

GB/T 11684-1989 堆用核仪器电磁干扰特性和检验方法

本标准主要涉及两种干扰耦合路径，即地和电源。 本标准给出了常见的干扰电平大小、单台仪器和系统安装前后抗干扰性能的检验方法，并规定了允许的界限值，作为单台仪器和系统安装的检验准则。 本标准适用于对电磁干扰非常敏感的、与安全有关的堆用核仪器。

Characterisics and test methods of electromagnetic interferences in nucleonic instrumentations for reactors

GB/T 9229-1988 放射性物质包装的内容物和辐射的泄漏检验

本标准规定了为运输放射性物质设计的货包原型的几种检验方法，这些方法是为了配合实施国家有关放射性物质安全运输规定标准而制订的，目的是为了检验经过上述运输规定中的试验后，货包中的放射性内容物是否仍无泄漏和货包外部辐射泄漏的增加是否仍在限量以下。 本标准并不能适用于所有的放射性物质货包，它的适用范围是： a.放射性内容物泄漏检验方法是针对包容放射性物质的某层包装或整个包容系统的密封性而制订的。它适用于低比活度的液体或粉末状物质的货包，如罐头盒和A型包装等。对装铀镭系放射性物质的货包，如果测量其子体（氡-222等）更为灵敏时，则不必采用本标准所阐述的方法，而可参考GB4075密封放射源分级中附录E1.1.5和E1.1.6方法进行。 b.辐射泄漏检验适用于屏蔽层外部的辐射剂量率，如果在设计中已考虑由于外层包装增加防护距离而使剂量率降低，那么也可连同外层包装检验，但对某些尺寸特殊或因某种特性使得检验过程困难的情况，不宜使用本标准的方法。

Radioactive materials--Packagings--Tests for contents leakage and radiation leakage

GB/T 9226-1988 标准放射源的检验证书

本标准规定了标准放射源检验证书的内容和基本要求。 本标准适用于各种标准放射性固体、液体和气体源。

Certification of standardized radioactive sources

GB/T 8997-1988 α.β表面污染测量仪与监测仪的校准

本标准适用于辐射防护领域中使用的可携式或固定式α、β和α-β表面污染测量仪与监测仪（测量的β最大能量大于0.15MeV），规定了对这类仪器的辐射特性进行定期或非定期校准的要求和方法。

Calibration for alpha,beta surface contamination meters and monitors

GB 8999-1988 电离辐射监测质量保证一般规定

本规定为制定电离辐射（以下简称辐射）监测质量保证计划提供一般原则，它适用于从事辐射防护工作的所有部门。

General rule of quality assurance for ionizing radiation monitoring

GB/T 8998-1988 环境热释光剂量计及其使用方法

本标准适用于环境热释光剂量计。规定了用于测量环境电离辐射空气吸收剂量的热释光剂量计的性能要求和使用方法。并推荐了性能检验方法等。

Environmental thermoluminescence dosimeter and its usage

GB/T 8996-1988 核电子仪器用样品盘尺寸

为了便于核电子仪器用样品盘的互换性，本标准规定了井型、平板和碟型样品盘尺寸，在特殊情况下，不排除使用其他尺寸或形状的样品盘。

Dimensions of sample planchets used in nuclear electronic instruments

GB/T 8995-1988 核反应堆中子注量率测量堆芯仪表

本标准适用于堆芯（在线）中子探测器及为安全、信息或控制目的所设计的仪表，也适用于反应堆主包壳内所包括的那些部件。常用的探测器是直流电离室、裂变电离室和自给能中子探测器。

In-core instrumentation for neutron fluence rate measurements in nuclear reactors

GB/T 8994-1988 辐射防护仪器的校准与定度x.Υ照射量率仪

本标准适用于在辐射防护领域中使用的携带式或固定式X、γ辐射照射量率仪、报警装置和监测仪。原则上也适用于那些实质上是用照射量校准和定度的辐射防护仪器。规定了对这一类仪器的辐射特性进行定期或非定期校准的要求和方法。 本标准不适用于反应堆控制装置和临界测量仪。

Calibration for radiation protection instruments X and gamma exposure rate meters

GB/T 5203-1985 核反应堆安全逻辑装置特性和检验方法

本标准等效采用国际标准IEC 744（1983）《核电站安全逻辑装置 特性和检验方法》。 本标准提出用于反应堆保护系统的安全逻辑装置设计、制造和检验的各项原则。 它应用GB 4083-83《核反应堆保护系统安全准则》提出的各项原则，并对这些原则的实现提出建议。 它包括验收和运行检验的规定、可靠性准则和消除外部各种影响的保护措施。 它不包括安全逻辑装置内部所完成的安全逻辑功能。

Safety logic assemblies of nuclear reactor characteristics and test methods

GB/T 12789.4-94 核反应堆仪表准则 第四部分: 液态金属冷却快堆

Nuclear Reactor Instrumentation Guidelines Part IV: Liquid Metal Cooled Fast Reactors

ASTM E2956-23 LWR反应堆压力容器中子暴露监测的标准指南

Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels

ASTM E521-23 使用带电粒子辐照研究中子辐射损伤影响的标准实施规程

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 chargedparticle 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.6 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 Practice for Investigating the Effects of Neutron Radiation Damage Using Charged-Particle Irradiation

ASTM E706-23 轻水反应堆压力容器监督标准主矩阵

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 responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.8 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 This master matrix 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 March 1, 2023. Published March 2023. Originally approved in 1979. Last previous edition approved in 2016 as E706 – 16. DOI: 10.1520/E0706-23. 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 2. Referenced Documents

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

ASTM D7013/D7013M-23 核表面压力计校准设施设置的标准指南

1.1 This guide outlines procedures for setup of a nuclear gauge calibration facility in either a shielded bay or an unshielded area—Guide A and Guide B, respectively. 1.2 This guide does not attempt to describe the calibration techniques or methods. It is assumed that this guide will be used by persons familiar with the operations of the gauge and in performing proper calibration, service and maintenance. 1.3 This guide does not attempt to address maintenance or service procedures related to the gauge. 1.4 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.6 This guide offers an organized collection of information or a series of options and does not recommend a specific course of action. This document cannot replace education or experience and should be used in conjunction with professional judgment. Not all aspects of this guide may be applicable in all circumstances. This ASTM standard is not intended to represent or replace the standard of care by which the adequacy of a given professional service must be judged, nor should this document be applied without consideration of a project’s many unique aspects. The word “Standard” in the title of this document has been approved through ASTM consensus process. 1.7 All observed and calculated values shall conform to the guidelines for significant digits and rounding established in practice D6026. 1.7.1 The method used to specify how data are collected, calculated, or recorded in this standard is not directly related to the accuracy to which the data can be applied in the design or other uses, or both. How one applies the results obtained using this standard is beyond its scope. 1.8 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 Calibration Facility Setup for Nuclear Surface Gauges

BS ISO 18077:2022 压水堆重装启动物理测试

1   Scope This document applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR. This document specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example, the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed).

Reload startup physics tests for pressurized water reactors

ISO 18077:2022 压水反应堆的重新加载启动物理试验

This document applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR. This document specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example, the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed).

Reload startup physics tests for pressurized water reactors

DB34/T 4255-2022 超导磁体制造过程管理规范

Superconducting magnet manufacturing process management specification

KS C IEC 60231-2022 核反应堆仪表的一般原理

General principles of nuclear reactor instrumentation