27.120.10 反应堆工程 标准查询与下载

EJ/T 20120-2016 钠冷快中子增殖堆设计准则 蒸汽发生器保护系统

本标准规定了钠冷快中子增殖堆蒸汽发生器保护系统设计的基本要求。 本标准适用于钠冷快中子增殖堆蒸汽发生器保护系统的设计。

Design criteria for sodium cooled fast breeder reactor.Steam generator protection system

EJ/T 20129-2016 钠冷快中子增殖堆操纵员培训及考试用模拟机

本标准规定了用于钠冷快中子增殖堆操纵员培训及考试的控制室模拟机功能要求本标准适用于钠冷快中子增殖堆控制室模拟机的验证确认、使用和软硬件配置管理。

Sodium cooled fast breeder reactor simulators for use in operator training and examination

EJ/T 20116-2016 钠冷快中子增殖堆设计准则 一回路冷却剂系统

本标准规定了池式钠冷快中子增殖堆一回路冷却剂系统的基本设计要求。 本标准适用于池式钠冷快中子增殖堆一回路冷却剂系统设计。

Design criteria for sodium cooled fast breeder reactor.Primary coolant system

EJ/T 20108-2016 钠冷快中子增殖堆设计准则 厂内辐射防护

本标准规定了钠冷快中子增殖堆厂房内辐射防护设计基本要求。 本标准适用于钠冷快中子增殖堆厂房内辐射防护设计。

Design criteria for sodium cooled fast breeder reactor.On-site radiation protection

EJ/T 20118-2016 钠冷快中子增殖堆设计准则 一回路钠净化系统

本标准规定了钠冷快中子增殖堆一回路钠净化系统的基本设计要求。 本标准适用于钠冷快中子增殖堆一回路钠净化系统的设计。

Design criteria for sodium cooled fast breeder reactor.Primary loop sodium purification system

ASTM E2006-16 轻水反应堆计算的基准试验的标准指南

3.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. The most important application of such calculations is the estimation of fluence within the reactor vessel of operating power plants to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, fission spectra are the fundamental data bases which control propagation of cross section uncertainties, while such physics-dosimetry experiments as vessel wall mockups, where measurements are made within a simulated reactor vessel wall, control error propagation associated with geometrical and methods approximations in the transport calculations. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response. 3.2 The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spectra and intensities. There are, however, less well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel”. When such mockups are suitably characterized they are also referred to as benchmark fields. A benchmark is that against which other things are referenced, hence the terminology “to benchmark reference” or “benchmark referencing”. A variety of benchmark neutron fields, other than standard neutron fields, have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. Some of these special benchmark experiments are discussed in this standard because they have identified needs for additional benchmarking or because they have been sufficiently documented to serve as benchmarks. 3.3 One dedicated effort to provide benchmarks whose radiation environments closely resemble those found outside the core of an operating reactor was the Nuclear Regulatory Commission's Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) (1)4. This program promoted better monitoring of the radiation exposure of reactor vessels and, thereby, provided for better assessment of vessel end-of-life conditions. An objective of the LWR-PV-SDIP was to develop improved procedures for reactor surveillance and document them in a series of ASTM standards (see Matrix E706). The primary means chosen for validating LWR-PV-SDIP procedures was by benchmarking a series of experimental and analytical studies in a variety of fields (see Guide E2005).

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

GSO ISO 7753:2015 核能临界检测和预警系统的性能和测试要求

  Scope is not provided for this standard

Nuclear energy -- Performance and testing requirements for criticality detection and alarm systems

GSO ISO 10270:2015 金属和合金的腐蚀核反应堆用锆合金的水腐蚀试验

This International Standard specifies: a) the determination of mass gain; b) the surface inspection of products of zirconium and its alloys when corrosion tested in water at 360 °C or in steam at or above 400 °C; c) that the tests in steam shall be performed at 10,3 MPa (1 500 psi). This International Standard is applicable to wrought products, castings, powder metallurgy products and weld metals. This method has been widely used in the development of new alloys, heat treating practices and for the evaluation of welding techniques, and should be utilized in its entirety to the extent specified for a product acceptance test, rather than merely a means of assessing performance in service.

Corrosion of metals and alloys -- Aqueous corrosion testing of zirconium alloys for use in nuclear power reactors

GSO IEC 60231:2015 核反应堆仪表的一般原理

1.2.1 This Recommendation gives guidance on the provision of reactor instrumentation and recommends standards of good practice. 1.2.2 The main body of the Recommendation is of general application and aspects applicable only to particular types of reactors are included in Appendices . 1.2.3 Items of instrumentation are included only where they have a direct bearing on the over-all safety and effective control of the reactor.

General principles of nuclear reactor instrumentation

NB/T 20360-2015 核电厂安全相关土建结构抗龙卷风设计规定

本标准规定了核电厂安全相关土建结构抗龙卷风设计的参数、荷载和荷载组合以及设计要求。 本标准适用于核电厂安全相关土建结构抗龙卷风设计。

Design Requirements for Tornado Protection of Nuclear Safety Related Civil Structures

NB/T 20359-2015 核电厂技术状态管理

本标准规定了核电厂技术状态管理的基本内容和基本要求。 本标准适用于核电厂构筑物、系统和部件(SSC)全寿期各阶段的技术状态管理。

Configuration management of nuclear power plants

NB/T 20361.1-2015 核电厂核岛机械设备在役试验 第一部分：通用要求

本部分规定了压水堆核电厂核岛机械设备在役试验的通用要求。 本部分适用于执行安全功能的泵、阈门、阻尼器的在役试验。

In-service test for mechanical conponents of NNP nuclear islands.Part 1: General requirements

NB/T 20361.2-2015 核电厂核岛机械设备在役试验 第二部分：泵

本部分规定了压水堆核电厂核岛内泵的在役试验要求。 本部分适用于安全相关的具有应急电源的离心泵和容积泵。 本部分不适用于下列设备： a)为运行方便单独提供应急电源的泵； b)作为其他部件的一部分由支架固定的泵，并经过充分试验验证。

In-service test for mechanical conponents of NNP nuclear islands.Part 2: Pump

NB/T 20361.3-2015 核电厂核岛机械设备在役试验 第三部分：阀门

本部分规定了压水堆核电厂核岛阀门的在役试验要求。 本部分适用于执行安全功能的阀门以及提供超压保护装置，但不包括下列阀门： a)仅用于为运行提供方便的阀门，例如排气、排水和试验阀门： b)仅用于系统控制的阀门，例如压力调节阀； c)仅用于系统或部件维修的阀门； d)移动式安装的阀门； e)负责检测各种电厂工况或为阀门运行提供信号的外部控制和保护系统； f)作为模块的一部分并参与整体试验，同时由营运单位确认已经充分试验的阀门。

In-service test for mechanical conponents of NNP nuclear islands.Part 2: Valve

ASTM D7013/D7013M-15 核表面测量仪校准设备的标准指南

5.1 To establish a proper calibration area for nuclear surface gauges. 5.2 To reduce the chance of improper calibration. Note 1: The quality of the results produced by this standard is dependent on the competence of the personnel performing it, and the suitability of the equipment and facilities used. Agencies that meet the criteria of practice D3740 are generally considered capable of competent and objective testing/inspection/etc. Users of this standard are cautioned that compliance with practice D3740 does not in itself assure a means of evaluating some of those factors. 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 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 guide to establish appropriate safety, and health 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 

Standard Guide for Calibration Facility Setup for Nuclear Surface Gauges

NB/T 20334-2015 压水堆核电厂反应堆及一回路噪声分析一般要求

本标准规定了压水堆核电厂反应堆及一回路噪声信号采集、分析、诊断及噪声分析程序执行的一般原则和基本技术要求。 本标准适用于压水堆核电厂反应堆及一回路噪声分析，其它堆型也可参照执行。

General requirements for noise analysis of reactor and primary loop in pressurized water reactor nuclear power plant

ASTM E900-15e2 用于预测反应堆容器材料中辐射诱导的转变温度变化的标准指南

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

ASTM E900-15e1 用于预测反应堆容器材料中辐射诱导的转变温度变化的标准指南

1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA).2,3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation. 1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation: 1.1.1.1 Copper content up to 0.4 %. 1.1.1.2 Nickel content up to 1.7 %. 1.1.1.3 Phosphorus content up to 0.03 %. 1.1.1.4 Manganese content within the range from 0.55 to 2.0 %. 1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F). 1.1.1.6 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV). 1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging). 1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation: 1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades. 1.1.2.2 Submerged arc welds, shielded arc welds, and electroslag welds having compositions consistent with those of the welds used to join the base materials described in 1.1.2.1. 1.1.2.3 Neutron fluence rate within the range from 3 × 1012 n/m2 /s to 5 × 1016 n/m2 /s (E > 1 MeV). 1.1.2.4 Neutron energy spectra within the range expected at the reactor vessel region adjacent to the core of commercial PWRs and BWRs (greater than approximately 500MW electric). 1.1.2.5 Irradiation exposure times of up to 25 years in boiling water reactors and 31 years in pressurized water reactors. 1.2 It is the responsibility of the user to show that the conditions of interest in their application of this guide are addressed adequately by the technical information on which the guide is based. It should be noted that the conditions quantified by the database are not distributed evenly over the range of materials and irradiation conditions described in 1.1, and that some combination of variables, particularly at the 1 This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.02 on Behavior and Use of Nuclear Structural Materials. Current edition approved Feb. 1, 2015. Published April 2015. Originally approved in 1983. Last previous edition approved in 2007 as E900 – 02(2007). DOI: 10.1520/E0900-15E01. 2 Available from ASTM International Headquarters. Order Adjunct No. ADJE090015-EA. 3 To inform the TTS prediction of Section 5 of this guide, the E10.02 Subcommittee decided to limit the data considered to Charpy shift values (∆T41J) measured from irradiations conducted in PWRs and BWRs. A database of 1,878 Charpy TTS measurements was compiled from surveillance reports on operating and decommissioned light water reactors of Western design from 13 countries (Brazil, Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea, Sweden, Switzlerland, Taiwan, and the United States), and from the technical literature. For each data record, the following information had to be available: fluence, fluence rate, irradiation temperature, and % content of Cu, Ni, P, and Mn. Reports and technical papers documenting the results of research programs conducted in material test reactors were also reviewed. Data from these sources was included in the database for information, but was not used in the development of the TTS prediction of Section 5 of this guide. 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 extremes of the data range are under-represented. Particular attention is warranted when the guide is applied to conditions near the extremes of the data range used to develop the TTS equation and when the application involves a region of the data space where data is sparse. Although the embrittlement correlation developed for this guide was based on statistical analysis of a large database, prudence is required for applications that involve variable values beyond the ranges specified in 1.1. Due to strong correlations with other exposure variables within the database (that is, fluence), and due to the uneven distribution of data within the database (for example, the irradiation temperature and flux range of PWR and BWR data show almost no overlap) neither neutron fluence rate nor irradiation time sufficiently improved the accuracy of the predictions to merit their use in the embrittlement correlation in this guide. Future versions of this guide may incorporate the effect of neutron fluence rate or irradiation time, or both, on TTS, as such effects are described in (1).4 The irradiated material database, the technical basis for developing the embrittlement correlation, and issues involved in its application, are discussed in a separate report (ADJE090015-EA). That report describes the nine different TTS equations considered in the development of this guide, some of which were developed using more limited datasets (for example, national program data (2, 3)). If the material variables or exposure conditions of a particular application fall within the range of one of these alternate correlations, it may provide more suitable guidance. 1.3 This guide is expected to be used in coordination with several standards addressing irradiation surveillance of lightwater reactor vessel materials. Method of determining the applicable fluence for use in this guide are addressed in Guides E482, E944, and Test Method E1005. The overall application of these separate guides and practices is described in Practice E853. 1.4 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to U.S. Customary units that are provided for information only and are not considered standard. 1.5 This standard guide does not define how the TTS should be used to determine the final adjusted reference temperature, which would typically include consideration of the transition temperature before irradiation, the predicted TTS, and the uncertainties in the shift estimation method. 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. 1.7 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 Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

ASTM E185-15 轻水慢化核电反应堆容器监督程序设计的标准实施规程

4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel. 4.2 The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guides E482 and E853. 4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects. 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 light-water small molecular 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 18201;×8201;1021 neutrons/m 2 (18201;×8201;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.......

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

ASTM C1533-15 热室设备的通用设计考量的标准指南

4.1 The purpose of this guide is to provide general guidelines for the design and operation of hot cell equipment to ensure longevity and reliability throughout the period of service. 4.2 It is intended that this guide record the general conditions and practices that experience has shown is necessary to minimize equipment failures and maximize the effectiveness and utility of hot cell equipment. It is also intended to alert designers to those features that are highly desirable for the selection of equipment that has proven reliable in high radiation environments. 4.3 This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for hot cell use. 4.4 This guide is intended to be generic and to apply to a wide range of types and configurations of hot cell equipment. 1.1 Intent: 1.1.1 The intent of this guide is to provide general design and operating considerations for the safe and dependable operation of remotely operated hot cell equipment. Hot cell equipment is hardware used to handle, process, or analyze nuclear or radioactive material in a shielded room. The equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. Therefore, the equipment is operated remotely, either with or without the aid of viewing. 1.1.2 This guide may apply to equipment in other radioactive remotely operated facilities such as suited entry repair areas, canyons or caves, but does not apply to equipment used in commercial power reactors. 1.1.3 This guide does not apply to equipment used in gloveboxes. 1.2 Applicability: 1.2.1 This guide is intended for persons who are tasked with the planning, design, procurement, fabrication, installation, or testing of equipment used in remote hot cell environments. 1.2.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded. 1.2.3 The system of units employed in this standard is the metric unit, also known as SI Units, which are commonly used for International Systems, and defined by IEEE/ASTM SI 10: American National Standard for Use of the International System of Units (SI): The Modern Metric System. 1.3 Caveats: 1.3.1 This guide does not ad......

Standard Guide for General Design Considerations for Hot Cell Equipment