F60 核反应堆综合 标准查询与下载

DIN 25493:2015 核设施.防止金属部件表面受到组装工具、密封、装璜、包装材料和隔热材料的损伤

Nuclear facilities - Protection of metallic surfaces of structural parts from damage from assembly aids, gaskets, packings, packaging material and thermal insulating materials

DIN 25459:2014 核电站用加筋和预应力混凝土外壳

Reinforced and prestressed concrete containment for nuclear power plants

KS C IEC 60557:2014 刷子架的定义和用语

Definition and terminology of brush--holders

NB/T 20297-2014 核电厂人员可靠性分析导则

本标准规定了核电厂人员可靠性分析与实施的基本要求,提供了人员可靠性分析的分析过程。本标准适用于核电厂人员可靠性分析。

Guide of human reliability analysis for nuclear power plants

ASME STP-NU-072-2014 小型模块化反应堆 (SMR) 路线图

This SMR Roadmap discusses Code areas in the ASME Boiler & Pressure Vessel Code (BPVC) and ASME Operation and Maintenance of Nuclear Power Plants Code (OM Code) where potential differences between vendors and NRC regarding the proper interpretation and application these Code requirements may present licensing issues. Particularly, this SMR Roadmap discusses potential issues in BPVC, Section III-Rules for Construction of Nuclear Facility Components (Section III), BPVC, Section XI-Rules for Inservice Inspection of Nuclear Power Plant Components (Section XI), and OM Code that may result from certain unique features of the SMR designs.

SMALL MODULAR REACTOR (SMR) ROADMAP

DIN 25474:2014 对于除反应堆外的核技术装备保持临界安全的管理措施

Measures of administrative character for conservation of criticality safety in nuclear facilities excluding reactors

IEEE 379-2014 核电站安全系统中单失效准则的应用

Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems

ASME STP-NU-069-2014 混凝土安全壳开裂评定用选定无损检测 (NDE) 方法分析

This report provides exploratory research into the reliability of available NDE technology to detect delaminations and cracking. It also provides the foundation for further interactions with the industry and ASME code to develop appropriate in-service inspection methodologies to perform detailed assessments of the integrity of aging concrete containment structures.

ANALYSIS OF SELECTED NONDESTRUCTIVE EXAMINATION (NDE) METHODOLOGIES FOR THE ASSESSMENT OF CRACKING IN CONCRETE CONTAINMENTS

ASME STP-NU-067-2014 聚变能量设备的ASME代码规则发展用路线图

This Roadmap has been developed as a guide to the Research & Development (R&D) and Code development tasks that could be considered in developing rules for Fusion Energy Devices (FED). The primary focus of the Roadmap is on the development of a complete set of Code rules for the design and operating conditions that are being proposed for the next generation fusion facilities.

ROADMAP FOR THE DEVELOPMENT OF ASME CODE RULES FOR FUSION ENERGY DEVICES

DIN 25435-6:2014 轻水反应堆初级冷却剂回路元件运行中检查.第6部分:蒸汽发生器加热管涡流检测

In-service inspections for primary coolant circuit components of light water reactors - Part 6: Eddy current testing of steam generator heating tubes

DIN 25435-2:2014 轻水反应堆初级冷却剂回路元件运行中检查.第2部分:磁粉检验和渗透检验

In-service inspections for primary coolant circuit components of light water reactors - Part 2: Magnetic particle and penetrant testing

DIN 25435-1:2014 轻水反应堆初级冷却剂回路元件运行中检查.第1部分:自动超声检查

In-service inspections for primary coolant circuit components of light water reactors - Part 1: Automated ultrasonic testing

DIN 25435-4:2014 轻水反应堆初级冷却剂回路元件运行中检查.第4部分:目视检查

In-service inspections for primary collant circuit components of light water reactors - Part 4: Visual testing

DIN 25435-7:2014 轻水反应堆初级冷却剂回路元件运行中检查.第7部分:放射性检测

In-service inspections for primary coolant circuit components of light water reactors - Part 7: Radiographic testing

ASTM E2956-14 监测LWR反应堆压力容器中子辐照的标准指南

4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 10-17 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time. 4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of the spatial core power distribution, including the time dependence. Because core power distributions change with time, reactor cavity or surveillance capsule measurements obtained early in plant life may not be representative of long-term reactor operation. Therefore, a simple normalization of neutron transport calculations to dosimetry data from a given capsule is unlikely to give a satisfactory solution to the problem over the full reactor lifetime. Guide E482 and Guide E944 provide detailed information related to the characterization of the neutron field for BWR and PWR power plants. 4.3 Fracture Mechanics Analysis—Currently, operating limitations for normal heat up and cool down transients imposed on the reactor pressure vessel are based on the fracture mechanics techniques outlined in the ASME Boiler and Pressure Vessel Code. This code requires the assumption of the presence of a surface flaw of depth equal to one fourth of the pressure vessel thickness. In addition, the fracture mechanics analysis of accident-induced transients (Pressurized Thermal Shock, (PTS)) may involve evaluating the effect of flaws of varying depth within......

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

ASTM E509/E509M-14 轻水慢化核反应堆容器运转中退火的标准指南

3.1 Reactor vessels made of ferritic steels are designed with the expectation of progressive changes in material properties resulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes in pressure-temperature (P – T) limits are made periodically during service life to account for the effects of neutron radiation on the ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictions on operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulated events, such as pressurized thermal shock (PTS). A reduction in the upper shelf toughness also occurs from neutron exposure, and this decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certain alternatives are available that reduce the problem or postpone the time at which plant restrictions must be considered. One of these alternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficiently above the normal operating temperature to recover a significant portion of the original fracture toughness and other material properties that were degraded as a result of neutron embrittlement. 3.2 Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guide the annealing operation. Sufficient time should be allocated to evaluate the expected benefits in operating life to be gained by annealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; to evaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as may be required to do the in-service annealing; and, to train personnel to perform the anneal. 3.3 Selection of the annealing temperature requires a balance of opposing conditions. Higher annealing temperatures, and longer annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the post-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the other hand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temper embrittlement. These higher temperatures also can cause engineering difficulties, that is, core and coolant removal and storage, localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports, primary coolant piping, adjacent concrete, insulation, etc. See ASME Code Case N-557 for further guidance on annealing conditions and thermal-stress evaluations (2). 3.3.1 When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is the number of years of additional service life that annealing of the vessel will provide. Two pieces of information are needed to answer the question: the post-anneal adjusted RTNDT and upper shelf energy level, and their subsequent changes during future irradiation. Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature limits for the period immediately following the anneal and demonstrating compliance wit...........

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

NB/T 20257-2013 核电厂操纵人员执照考核

本标准规定了核电厂操纵人员申请执照、换发执照和执照转移考核活动的相关内容以及知识技能的考核要求,主要包括:a) 组织机构与职责;b) 参加考核人员的资格要求;c) 考核管理与流程;d) 取照考试、换发新执照考试及执照转移考试中,模拟机考试、笔试、口试、现场考试等4个部分的考核实施要求和评定标准。本标准适用于中华人民共和国核电厂操纵人员申请执照、换发执照和执照转移的考核与管理工作。

Nuclear power plant operators' license examination

CFR 10-100.20-2013 能源. 第100部分:反应堆厂址标准. 第100.20:评估站点时需要考虑的因素.

The Commission will take the following factors into consideration in determining the acceptability of a site for a stationary power reactor.

Energy.Part100:Reactor site criteria. Section100.20:Factors to be considered when evaluating sites.

CFR 10-100.21-2013 能源. 第100部分:反应堆厂址标准. 第100.21:非地震选址标准.

Applications for site approval for commercial power reactors shall demonstrate that the proposed site meets the following criteria.

Energy.Part100:Reactor site criteria. Section100.21:Non-seismic siting criteria.

EJ/T 737-2012 压水堆噪声分析一般要求

General Requirements for PWR Noise Analysis