F69 核反应堆与核电厂核岛设备 标准查询与下载

ASME N-2-2011 相同核部件及附件合格证书持有人的数据报告

Certificate Holders' Data Report For Identical Nuclear Parts and Appurtenances

ASME N-1-2011 核反应堆容器合格证书持有人的数据报告

Certificate Holders' Data Report For Nuclear Vessels

ASME NPP-1-2011 预制的核管道组件合格证书持有人的数据报告

Certificate Holders' Data Report For Fabricated Nuclear Piping Sub-Assemblies

ASME N-5-2011 核电站部件,支架及附属物合格证书持有人的安装或工厂装配数据报告

Certificate Holders' Data Report for Installation or Shop Assembly of Nuclear Power Plant Components, Supports, and Appurtenances

ASME C-1-2011 混凝土反应堆容器和安全壳合格证书持有人的数据报告

Certificate Holders' Data Report For Concrete Reactor Vessels and Containments

ASME NQA-1B-2011 附录ASME NQA-1–2008质量保证要求核设施应用

Addenda to ASME NQA-1–2008 Quality Assurance Requirements for Nuclear Facility Applications

IEEE 628-2011 核电站1E级电路的电缆管道系统的设计、安装和鉴定

Criteria for the Design, Installation, and Qualification of Raceway Systems for Class 1E Circuits for Nuclear Power Generating Stations

ASME STP-NU-041-2011 更新和改进分段NH.可替代简化的蠕变疲劳设计方法

This report investigates five newly proposed creep-fatigue evaluation methods to improve the provisions on creep-fatigue evaluation of Mod.9Cr-1Mo steel in ASME Subsection NH (Class 1 Components in Elevated Temperature Service) of ASME Boiler and Pressure Vessel Code Section III: 1) Modified Ductility Exhaustion Method (MDEM), 2) Strain Range Separation Method (SRSM), 3) Approach for Pressure Vessel Applications (APVA), 4) Hybrid Method of Time Fraction and Ductility Exhaustion (Hybrid) 5) Simplified Model Test Approach (SMT).

UPDATE AND IMPROVE SUBSECTION NH �C ALTERNATIVE SIMPLIFIED CREEP-FATIGUE DESIGN METHODS

ANSI/ANS 53.1-2011 模块化氢冷堆核电厂核安全的设计流程

Applies to the safety design process for MHR nuclear power plants. This standard provides a process for establishing top-level safety criteria (TLSC), safety functions, top-level design criteria (TLDC), licensing basis events (LBEs), design basis accidents (DBAs), safety classification of systems, structures, and components (SSC), safety analyses, defense-in-depth (DID), and adequate assurance of special treatment requirements for safety-related SSC throughout the operating life of the plant. The standard does not provide detailed guidance for design; other existing standards cover those.

Nuclear Safety Design Process for Modular Helium-Cooled Reactor Plants

ASME STP-NU-044-2011 高温反应堆的非破坏性测试(NDE)和在役检测技术(ISI)

This report identifies several Non Destructive Examination (NDE) technologies of advanced monitoring, diagnostics and prognostics systems (e.g., acoustic emission, ultrasonic, advanced material characterization) applicable to components of High Temperature Gas-cooled Reactors (HTGRs) for in-service inspection.

NON DESTRUCTIVE EXAMINATION (NDE) AND IN-SERVICE INSPECTION (ISI) TECHNOLOGY FOR HIGH TEMPERATURE REACTORS

ASTM E1214-11(2018) 反应堆容器监视用熔丝温度监测器的标准使用指南

3.1 Temperature monitors are used in surveillance capsules in accordance with Practice E2215 to estimate the maximum value of the surveillance specimen irradiation temperature. Temperature monitors are needed to give evidence of overheating of surveillance specimens beyond the expected temperature. Because overheating causes a reduction in the amount of neutron radiation damage to the surveillance specimens, this overheating could result in a change in the measured properties of the surveillance specimens that would lead to an unconservative prediction of damage to the reactor vessel material. 3.2 The magnitude of the reduction of radiation damage with overheating depends on the composition of the material and time at temperature. Guide E900 provides an accepted method for quantifying the temperature effect. Because the evidence from melt wire monitors gives no indication of the duration of overheating above the expected temperature as indicated by melting of the monitor, the significance of overheating events cannot be quantified on the basis of temperature monitors alone. Indication of overheating does serve to alert the user of the data to further evaluate the irradiation temperature exposure history of the surveillance capsule. 3.3 This guide is included in Master Matrix E706 that relates several standards used for irradiation surveillance of light water reactor vessel materials. It is intended primarily to amplify the requirements of Practice E185 in the design of temperature monitors for the surveillance program. It may also be used in conjunction with Practice E2215 to evaluate the post-irradiation test measurements.. 1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practices E185 and E2215. 1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties. 1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to inch-pound units that are provided for information only and are not considered standard. 1.4 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. (See Note 1.)

Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance

ANSI/ANS 2.3-2011 评估核设施场地的龙卷风,飓风和极端直线风特性

This standard defines site phenomena caused by (1) extreme straight winds, (2) hurricanes, and (3) tornadoes in various geographic regions of the U.S. These phenomena are used for the design of nuclear facilities.

Estimating Tornado, Hurricane, and Extreme Straight Line Wind Characteristics at Nuclear Facility Sites

ASME STP-NU-042-2011 ASME分段NH的新材料

New Materials for ASME Subsection NH

DIN EN 62340:2010 核电站.仪器和控制设备对安全的重要性.应对共因失效(CCF)的要求(IEC 62340-2007).德文版本EN 62340-2010

Nuclear power plants - Instrumentation and control systems important to safety - Requirements for coping with Common Cause Failure (CCF) (IEC 62340:2007); German version EN 62340:2010

BS ISO 26802:2010 核设施.核反应堆用防泄漏和通风系统设计和操作准则

Nuclear facilities. Criteria for the design and the operation of containment and ventilation systems for nuclear reactors

DIN EN 61226:2010 核电站.仪器和控制对安全的重要性.安全系统测量途径的性能评估用方法(IEC 61226-2009).德文版本EN 61226-2010

Nuclear power plants - Instrumentation and control important to safety - Classification of instrumentation and control functions (IEC 61226:2009); German version EN 61226:2010

ISO 26802:2010 核设施.核反应堆用防泄漏和通风系统的设计和操作准则

his International Standard specifies the applicable requirements related to the design and the operation of containment and ventilation systems of nuclear power plants and research reactors, taking into account the following.

Nuclear facilities - Criteria for the design and the operation of containment and ventilation systems for nuclear reactors

DIN EN 60964:2010 核电站.控制室.设计(IEC 60964-2009).德文版本EN 60964-2010

Nuclear power plants - Control rooms - Design (IEC 60964:2009); German version EN 60964:2010

BS IEC 62465:2010 核电站.仪器使用和控制安全性重要事项.电力电缆系统老化管理

Nuclear power plants - Instrumentation and control important to safety - Management of ageing of electrical cabling systems

ASME BPVC Section 3 Division 1 Appendices-2010 ASME锅炉和压力容器规范.第3节:核设备组件的制造规则.第1分册:附录

ASME Boiler & Pressure Vessel Code - Section 3: Rules for Construction of Nuclear Facility Components - Division 1: Appendices