F60 核反应堆综合 标准查询与下载

ASTM F2789-10 核装置的机械和功能特征表示的标准指南

Nucleus devices are generally designed to augment the mechanical function of native degenerated nucleus material or to replace tissue that has been removed during a surgical procedure. This guide outlines methods for evaluating many different types of devices. Comparisons between devices must be made cautiously and with careful analysis, taking into account the effects that design and functional differences can have on the testing configurations and overall performance, and the possibility that mechanical failure may not be related to clinical failure and inversely, that mechanical success may not be related to clinical success. These tests are conducted in vitro to allow for analysis of the mechanical performance of the nucleus device under specific testing modalities. The loads applied may differ from the complex loading seen in vivo, and therefore the results from these tests may not directly predict in vivo performance. These tests are used to quantify the static and dynamic properties and performance of different implant designs. The mechanical tests are conducted in vitro using simplified loads and moments. Fatigue testing in a simulated body fluid or saline may have fretting, aging, corroding, or lubricating effects on the device and thereby affect the relative performance of tested devices. Hence, the test environment and the effect of that environment, whether a simulated body fluid, normal saline bath (9 g NaCl per 1000 mL H2O), or dry, is an important characteristic of the test and must be reported accurately. Dynamic testing methods should be designed to answer the following questions, including but not limited to: Does the device still function as intended after cycling? Does it retain adequate performance characteristics (for example, mechanical and kinematic properties such as ROM)? Did the device wear or degrade? If there is evidence of wear or degradation of the device, it should be identified and quantified with reasonable methods generally available. The user shall distinguish between particulates generated by the device and particulates generated by the test model and fixtures if technically feasible.1.1 This guide describes various forms of nucleus replacement and nucleus augmentation devices. It further outlines the types of testing that are recommended in evaluating the performance of these devices. 1.2 Biocompatibility of the materials used in a nucleus replacement device is not addressed in this guide. However, users should investigate the biocompatibility of their device separately (see X1.1). 1.3 While it is understood that expulsion and endplate fractures represent documented clinical failures, this guide does not specifically address them, although some of the factors that relate to expulsion have been included (see X1.3). 1.4 Multiple tests are described in this guide; however, the user need not use them all. It is the responsibility of the user of this guide to determine which tests are appropriate for the devices being tested and their potential application. Some tests may not be applicable for all types of devices. Moreover, some nucleus devices may not be stable in all test configurations. However, this does not necessarily mean that the test methods described are unsuitable. 1.5 The science of nucleus device design is still very young and includes technology that is changing more quickly than this guide can be modified. Therefore, the user must carefully consider the applicability of this guide to the user’s particular device; the guide may not be appropriate for every device. For example, at the time of publication, this guide does not address the nucleus replacement and nucleus augmentation devi......

Standard Guide for Mechanical and Functional Characterization of Nucleus Devices

ASME STP-NU-037-2010 美国机械工程师学会(ASME)第Ⅲ部分第NH分段中操作条件容许应力值

This report provides an evaluation of the consistency of the stress allowable in ASME Subsection NH - Class 1 Components in Elevated Temperature Service (ASME III-NH) with the ASME Section II Part D (ASME II-D) values for each of the materials currently covered by ASME III-NH and reviews the availability of the original and augmented databases for the materials. The report also analyzes the expanded databases and recommends actions to meet the needs for setting allowables over the range of temperatures and times of interest to the Generation IV reactor concepts.

OPERATING CONDITION ALLOWABLE STRESS VALUES IN ASME SECTION III SUBSECTION NH

ANSI/IEEE 7-4.3.2:2010 核动力发电站安全系统中数字计算机的标准

This standard serves to amplify criteria in IEEE Std 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations, to address the use of computers as part of safety systems in nuclear power generating stations. The criteria contained herein, in conjunction with criteria in IEEE Std 603-2009, establish minimum functional and design requirements for computers used as components of a safety system.

Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations

ASTM D5411-10 反应堆冷却剂中放射性核素混合物每次蜕变的平均能量(E)计算的标准操作规程

This practice is useful for the determination of the average energy per disintegration of the isotopic mixture found in the reactor-coolant system of a nuclear reactor (1).1.1 This practice applies to the calculation of the average energy per disintegration (E–) for a mixture of radionuclides in reactor coolant water. 1.2 The microcurie (µCi) is the standard unit of measurement for this standard. The values given in parentheses are mathematical conversions to SI units, which are provided for information only and are not considered standard. 1.3 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 Calculation of Average Energy Per Disintegration (E) for a Mixture of Radionuclides in Reactor Coolant

ASTM C1725-10 热室专用辅助设备和工具标准指南

This guide is relevant to the design of specialized support equipment and tools that are remotely operated, maintained, or viewed through shielding windows, or combinations thereof, or by other remote viewing systems. Hot cells contain substances and processes that may be extremely hazardous to personnel or the external environment, or both. Process safety and reliability are improved with successful design, installation, and operation of specialized mechanical and support equipment. Use of this guide in the design of specialized mechanical and support equipment can reduce costs, improve productivity, reduce failed hardware replacement time, and provide a standardized design approach.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.

Standard Guide for Hot Cell Specialized Support Equipment and Tools

ANSI/NFPA 805-2010 轻水反应堆发电站的防火安全基本性能标准

1.1 Scope. This standard specifies the minimum fire protection requirements for existing light water nuclear power plants during all phases of plant operation, including shutdown, degraded conditions, and decommissioning.

Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants

ASTM E2215-10 轻水中型核动力反应堆罐监测室评估的标准实施规程

Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules. Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636. This standard practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185. The radiation-induced changes in the properties of the vessel are generally monitored by measuring the Charpy index temperatures, the Charpy upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185. The application of this data is the subject of Guide E900 and other documents listed in Section 2. Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide E636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E1921 or J-integral techniques defined in Test Method E1820. Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the radiation response of the materials. The methodology to be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence is defined in Practice E853. Guide E900 describes the bases used to evaluate the radiation-induced changes in Charpy transition temperature for reactor vessel materials and provides......

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

ASTM E2006-10 轻水反应堆计算水准试验标准指南

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. 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. 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) . 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).1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

KS C IEC 61788-1:2009 超导性.第1部分:临界电流测量-Cu/Nb-Ti复合超导体的直流临界电流

이 표준은 구리와 초전도체의 비가 1 이상인 Cu/Nb-Ti 복합 초전도체의 직류 임계전류

Superconductivity－Part 1：Critical current measurement－DC critical current of Cu/Nb-Ti composite superconductors

IEC 61500:2009 核电站.重要安全的仪表与控制系统.系统执行A类函数的数据通信

This International Standard establishes requirements for data communication which is used in systems performing category A functions in nuclear power plants. It covers also interface requirements for data communication of equipment performing category A functions with other systems including those performing category B and C functions and functions not important to safety. The scope of this standard is restricted to the consideration of data communication within the plant IC systems. It does not cover communication by telephone@ radio@ voice@ fax@ email@ public address etc. The internal operation and the detailed technical specification of data communication equipment are not in the scope of this standard. This standard is not applicable to the internal connections and data communication of a processor unit@ its memory and control logic. It does not concern the internal processing of instrumentation and control computer systems. This standard gives requirements for functions and properties of on-line plant data communications by reference to IEC 60880 and IEC 60987@ produced within the framework of IEC 61513. It requires classification of the communication functions in accordance with IEC 61226@ which in turn requires environmental and seismic qualification (i.e.@ the environment where the safety function is required to operate) according to IEC 60780 and IEC 60980

Nuclear power plants - Instrumentation and control important to safety - Data communication in systems performing category A functions

IEC 60964:2009 核电站.控制室.设计

Scope and object This International Standard establishes requirements for the human-machine interface in the main control rooms of nuclear power plants. The standard also establishes requirements for the selection of functions@ design consideration and organization of the human-machine interface and procedures which shall be used systematically to verify and validate the functional design. These requirements reflect the application of human factors engineering principles as they apply to the human-machine interface during normal and abnormal plant conditions. This standard does not cover special purpose or normally unattended control points@ such as those provided for shutdown operations from outside the main control room or for radioactive waste handling@ or emergency response facilities. Detailed equipment design is outside the scope of this standard. The primary purpose of this standard is to provide functional design requirements to be used in the design of the main control room of a nuclear power plant to meet operational and safety requirements. This standard also provides functional interface requirements which relate to control room staffing@ operating procedures@ and the training programmes which@ together with the human-machine interface@ constitute the control room system. This standard is intended for application to new control rooms whose conceptual design is initiated after the publication of this standard. If it is desired to apply it to an existing control room@ special caution must be exercised so that the design basis is kept consistent.

Nuclear power plants - Control rooms - Design

IEEE 535/INT 1-2009 IEEE Std 535 ?-2006的IEEE标准解释.核动力发电站用1E级铅蓄电池的鉴定用IEEE标准

This is an interpretation of IEEE Std 535-2006. Interpretations are issued to explain and clarify the intent of a standard and are not intended to constitute an alteration to the original standard or to supply consulting information. Permission is hereby granted to download and print one copy of this document. Individuals seeking permission to reproduce and/or distribute this document in its entirety or portions of this document must contact the IEEE Standards Department for the appropriate license. Use of the information contained in this document is at your own risk.

IEEE Standards Interpretations for IEEE Std 535 ?-2006 IEEE Standard for Qualification of Class 1E Lead Storage Batteries for Nuclear Power Generating Stations

ASME RA-Sa Addenda-2009 核电厂设施的概率风险评估.勘误表

This document sets forth the requirements for probabilistic risk assessments (PRAs) used to support riskinformed decisions for commercial light water reactor nuclear power plants and prescribes a method for applying these requirements for specific applications.

Probalistic risk assessment for nuclear power plant applications; Addenda

ASME RA-Sa Errata-2009 核电站设施用1级/大型早期释放频率可能性风险评估标准.勘误表

Standard for Level 1/Large Early Release Frequency - Probalistic risk assessment for nuclear power plant applications; Errata

ANSI/ANS 19.10-2009 沸水堆(BWR)和压水反应堆(PWR)压力容器及反应器内部中中子注量的测定方法

This standard provides criteria for performing and validating the sequence of calculations required for the prediction of the fast neutron fluence t in the reactor vessel. Applicable to PWR and BWR plants the standard addresses flux attenuation from the core through the vessel to the cavity and provides criteria for generating cross sections, spectra, transport and comparisons with in- and ex-vessel measurements, validation, uncertainties and flux extrapolation to the inside vessel surface.

Methods for Determining Neutron Fluence in BWR and PWR Pressure Vessel and Reactor Internals

BS IEC 62340:2008 核电站.仪器和控制设备的安全重要性.应对共因失效(CCF)的要求

I&C systems important to safety may be designed using conventional hard-wired equipment, computer-based equipment or by using a combination of both types of equipment. This International Standard provides requirements and recommendations1 for the overall architecture of I&C systems, which may contain either or both technologies. The scope of this standard is: a) to give requirements related to the avoidance of CCF of I&C systems that perform category A functions; b) to additionally require the implementation of independent I&C systems to overcome CCF, while the likelihood of CCF is reduced by strictly applying the overall safety principles of IEC SC 45A (notably IEC 61226, IEC 61513, IEC 60880 and IEC 60709); c) to give an overview of the complete scope of requirements relevant to CCF, but not to overlap with fields already addressed in other standards. These are referenced. This standard emphasises the need for the complete and precise specification of the safety functions, based on the analysis of design basis accidents and consideration of the main plant safety goals. This specification is the pre-requisite for generating a comprehensive set of detailed requirements for the design of I&C systems to overcome CCF. This standard provides principles and requirements to overcome CCF by means which ensure independence2: a) between I&C systems performing diverse safety functions within category A which contribute to the same safety target; b) between I&C systems performing different functions from different categories if e.g. a category B function is claimed as back-up of a category A function and; c) between redundant channels of the same I&C system. The implementation of these requirements leads to various types of defence against initiating CCF events. Means to achieve protection against CCF are discussed in this standard in relation to: a) susceptibility to internal plant hazards and external hazards; b) propagation of physical effects in the hardware (e.g. high voltages); and c) avoidance of specific faults and vulnerabilities within the I&C systems notably: 1) propagation of functional failure in I&C systems or between different I&C systems (e.g. by means of communication, fault or error on shared resources), 2) existence of common faults introduced during design or during system operation (e.g. maintenance induced faults), 3) insufficient system validation so that the system behaviour in response to input signal transients does not adequately correspond to the intended safety functions, 4) insufficient qualification of the required properties of hardware, insufficient verification of software components, or insufficient verification of compatibility between replaced and existing system components.

Nuclear power plants - Instrumentation and control systems important to safety - Requirements for coping with common cause failure (CCF)

ANSI/ASTM C781:2008 高温冷却气体核反应堆用石墨和硼化石墨部件测试用实施规程

Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors

ASTM D4286-08 核动力发电机装置用涂覆承包人资格鉴定的标准实施规程

1.1 This practice provides a criteria guide and procedural method to assist utility owners, architects, engineers, constructors, and other selection agencies in determining the overall qualifications of a coating contractor to execute coating work for the primary containment and other safety-related facilities of light-water nuclear power plants. 1.2 The qualification criteria and requirements address the essential basic capability of a contractor to execute nuclear coating work. Obviously, the specific capability to execute those requirements unique to a given project must also be carefully considered. The evaluation procedure contained in this practice is designed to be adaptive to this detailed final qualification process. Variation or simplification of the practice is appropriate for non safety-related areas of nuclear power plants, fossil fueled facilities, and other industrial projects. 1.3 The overall capability of a contractor to successfully execute the varied and complex requirements of nuclear coating work is dependent upon competency in a variety of essential categories. 1.4 The nine evaluation categories described in Sections 3-12 detail the specific data to be provided by the contractor as essential to determining qualification status. In addition, a tenth untitled category has been provided on the evaluation work described in Section 12 for inclusion of other information pertinent to a specific project. 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 Determining Coating Contractor Qualifications for Nuclear Powered Electric Generation Facilities

ASTM C781-08(2014) 高温气体冷却核反应堆部件用石墨和硼化石墨材料测试用标准实施规程

4.1 Property data obtained with the recommended test methods identified herein may be used for research and development, design, manufacturing control, specifications, performance evaluation, and regulatory statutes pertaining to high temperature gas-cooled reactors. 4.2 The test methods are applicable primarily to specimens in the non-irradiated and non-oxidized state. Many are also applicable to specimens in the irradiated or oxidized state, or both, provided the specimens meet all requirements of the test method. The user is cautioned to consider the instructions given in the test methods. 4.3 Additional test methods are in preparation and will be incorporated. The user is cautioned to employ the latest revision. 1.1 This practice covers the test methods for measuring the properties of graphite and boronated graphite materials. These properties may be used for the design and evaluation of high-temperature gas-cooled reactor components. 1.2 The test methods referenced herein are applicable to materials used for replaceable and permanent components as defined in Section 7 and Section 9, and includes fuel elements; removable reflector elements and blocks; permanent side reflector elements and blocks; core support pedestals and elements; control rod, reserve shutdown, and burnable poison compacts; and neutron shield material. 1.3 This practice includes test methods that have been selected from existing ASTM standards, ASTM standards that have been modified, and new ASTM standards that are specific to the testing of materials listed in 1.2. Comments on individual test methods for graphite and boronated graphite components are given in Sections 8 and 10, respectively. The test methods are summarized in Tables 1 and 2. TABLE 1 Summary of Test Methods for Graphite Components Note 1: Designations under preparation will be added when approved.  

Standard Practice for Testing Graphite and Boronated Graphite Materials for High-Temperature Gas-Cooled Nuclear Reactor Components

ANSI/ANS 8.27-2008 轻水反应堆(LWR)燃料用燃耗信贷

The standard provides criteria for process and techniques used for criticality safety evaluations of irradiated light water reactor fuel assemblies in storage, transportation and disposal.

Burnup Credit for LWR Fuel