F60 核反应堆综合 标准查询与下载

ASME N-8-2002 SC或TC级外壳焊接数据报告格式

Class SC or Tc Closure Welds Data Report Forms (07/01)

Class SC or Tc Closure Welds Data Report Forms (07/01)

ASME N-9-2002 厂制外壳部件和附件数据报告

Shop Fabricated Containment Parts and Appurtenances Data Report (07/01)

Shop Fabricated Containment Parts and Appurtenances Data Report (07/01)

ASME N-7-2002 核反应堆外壳数据报告

Nuclear Containments Data Report (07/01)

Nuclear Containments Data Report (07/01)

ASTM E185-02 轻水冷却核反应堆容器的监督程序设计的标准操作规程

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 tests yield results useful for the evaluation of radiation effects on the reactor vessel. The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guide E 482, which establishes the bases to be used to evaluate both the design and future condition of the reactor vessel. 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 the beltline of light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and a 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) at the end of the design lifetime (EOL) exceeds 1 x 1017 n/cm2 (1 x 1021 n/m2) at the inside surface of the reactor vessel.1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E 185 apply to earlier reactor vessels.1.4 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life, but the procedure described may provide guidance for developing such a surveillance program.Note 18212;The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E 185 describes the minimum requirements for a surveillance program. Practice E 2215, "Standard Practice for the Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels" describes the procedures for testing and evaluation of surveillance capsules removed from a surveillance program as defined in the current or previous editions of Practice E 185. Another standard guide for supplementing existing light-water moderated nuclear power reactor vessel surveillance programs is under preparation. A summary of the many major revisions to Practice E 185 since its original issuance is contained in Appendix X1.

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

ANSI/ANS 58.9-2002 轻水反应堆有关安全的液体系统用单一故障标准

Provides criteria for the designer which interpret the requirements of Title 10, Code of Federal Regulations, Part 50, "Licensing of Production and Utilization Facilities, " Appendix A, "General Design Criteria for Nuclear Power Plants, " with respect to design against single failures in safety-related Light Water Reactor (LWR) fluid systems [1]. Means of treating both active and passive failures are addressed for safety-related fluid systems following various initiating events. Current acceptable practice is used as a basis for these criteria. Failure criteria for the electric power systems and the protection systems are provided in IEEE Std 308-1980 "IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations", IEEE Std 279-1971 "IEEE Standard Criteria for Protection Systems for Nuclear Power Generating Stations" (N42.7-1971), IEEE Std 379-1977 "IEEE Standard for Application of the Single-Failure Criterion to Nuclear Power Generating Station Class IE Systems", and IEEE Std 603-1980 "Standard Criteria for Safety Systems for Nuclear Power Generating Stations." {2, 3, 4, 5}. Failures of structural components, such as braces, supports, or restraints, as well as occurrences involving common mode failures, are excluded.

Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems

ASTM E2216-02(2013) 核设施中退役混凝土处理选项评估的标准指南

4.1 This standard guide applies to concrete that is still in place with a defined geometry and known, documented history. 4.2 It is not intended for use on concrete that has already been rubbelized where it is difficult to measure the radiation levels and not easy to remove surface contamination to reduce radiation levels after concrete has been rubbelized. 4.3 This standard guide applies to surface or volumetrically contaminated concrete, where the depth of contamination can be measured or estimated based on the history of the concrete. 4.4 This standard guide does not apply to the reinforcement bar (rebar) found in concrete. Although most concrete contains rebar, it is generally removed before the concrete is dispositioned. In addition, rebar may be activated, and is covered under procedures for reuse of scrap metal. 4.5 General unit-dose and unit-cost data to support the calculations is provided in the appendices of this standard guide. However, if site-specific data is available, it should be used instead of the general information provided here. 4.6 This standard guide helps determine estimated doses to the public during disposal of concrete and to future residents of disposal areas. It does not include dose to radiation workers already involved in a radiation control program. It is assumed that the dose to radiation workers is already tracked and kept within acceptable levels through a radiation control program. The cost and dose to radiation workers could be added in to find an overall cost and dose for each option. 1.1 This standard guide defines the process for developing a strategy for dispositioning concrete from nuclear facility decommissioning. It outlines a 10-step method to evaluate disposal options for radioactively contaminated concrete. One of the steps is to complete a detailed analysis of the cost and dose to nonradiation workers (the public); the methodology and supporting data to perform this analysis are detailed in the appendices. The resulting data can be used to balance dose and cost and select the best disposal option. These data, which establish a technical basis to apply to release the concrete, can be used in several ways: (1) to show that the release meets existing release criteria, (2) to establish a basis to request release of the concrete on a case-by-case basis, (3) to develop a basis for establishing release criteria where none exists. 1.2 This standard guide is based on the ???Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Sites,??? (1)2 from which the analysis methodology and supporting data are taken. 1.3 Guide E1760 provides a general process for release of materials containing residual amounts of radioactivity. In addition, Guide

Standard Guide for Evaluating Disposal Options for Concrete from Nuclear Facility Decommissioning

DB65/ 009-2001 发供电系统电气操作票;工作票考核标准

Examine standard for the electricity operating card and working card in the electricity generating and electrcity supplying system

DB65/ 010-2001 发电厂热力系统操作票;工作票考核标准

Examine standard for operating card and working card in power plant thermal system

DIN IEC 62117:2001 核电厂.压水堆(PWR).低温停机过程中堆芯充分冷却监测

Nuclear power plants - Pressurized light water reactors - Monitoring adequate cooling within the core during cold shutdown (IEC 62117:1999)

ISO 15080:2001 核设施 屏蔽罩的通风

This International Standard specifies the requirements for the construction and the installation of radiobiological shielding devices used as ventilation passages through shielded enclosures with concrete or leaded walls to protect against gamma radiation. This International Standard applies to all shielded containment enclosures used for handling radioactive products or material emitting penetrating radiation (gamma or neutrons) in such quantities and of such emission rate that these products must be handled remotely behind a shielding wall. Typically, the enclosures considered cover all types of nuclear fuel cycle installations: reprocessing plants, hot activity laboratories, plutonium solution handling facilities, shielded cells, waste storage installations, etc. It could eventually be applied to particle accelerators, primary containment of research reactors, fusion research reactors, radiographic installations, neutron generators, etc. However, pressurized vessels, sealed sources, transport packaging for radioactive materials, as well as enclosures, primary circuits and vessels of nuclear power plants have been deliberately excluded from the scope of this International Standard. This International Standard specifies general and detailed principles which shall be respected when designing ventilation penetrations for shielded enclosures. These specifications can be divided more generally into two categories of guidance, which apply to the two following systems of ventilation penetrations for shielded enclosures already in use: — the first corresponding to the most important conventional systems used worldwide, and — the second corresponding to an alternative method, called the "cast iron helix technique".

Nuclear facilities - Ventilation penetrations for shielded enclosures

BS IEC 61501:2001 核反应堆仪器仪表.宽范围中子流量仪.平方电压法

This International Standard applies to instrument and measurement channels which generate a calculation of the mean square voltage (MSV) of a signal arising from a neutron detector, in order to extract from it information relating to the neutron fluence rate of a nuclear reactor. After calibration, this information can be used to derive the relative power and the time constant, for example expressed in terms of period, doubling time, decades per minute or percent per second. The method used to calculate the mean square voltage of the signal is also known as "fluctuation treatment" or "the Campbell method". Associated with other techniques of measurement, such as pulse rate counting or current measurement, the calculation of the mean square voltage allows the assembly of a series of wide range neutron fluence rate measurements for the simplification of nuclear instrumentation systems in the control of nuclear reactors. This standard describes the principles, the terminology, the characteristics, the requirements and the testing methods related to instrumentation and measurement of the neutron fluence rate using MSV techniques for nuclear reactor control. Typical examples of the application of the MSV techniques are given.

Nuclear reactor instrumentation - Wide range neutron fluence rate meter - Mean square voltage method

NF M60-315:2001 核能.核燃料技术.废料.用无源中子计数法对放射性废料形态活性测量装置校准的推荐规范

Nuclear energy - Nuclear fuel technology - Waste - Recommendations for the calibration of an activity measurement facility of radioactive waste forms by passive neutron counting.

NF M60-317:2001 核能.燃料循环技术.废物.预先化学分离后通过液体闪烁测定废水和废物中的镍63

Énergie nucléaire - Technologie du cycle du combustible - Déchets - Détermination du nickel 63 dans les effluents et déchets par scintillation liquide, après séparation chimique préalable

IEC 61513:2001 核电厂 对安全重要的系统的仪器仪表和控制设备 系统的一般要求

1 General I&C systems important to safety may be implemented using conventional hard-wired equip-ment, computer-based (CB) equipment or by using a combination of both types of equipment. This International Standard provides requirements and recommendations (see note) for the total I&C system architecture which may contain either or both technologies. NOTE In the following, the term requirements is used as a comprehensive term for both requirements and recommendations. The distinction appears at the level of the specific provisions, where requirements are expressed by "shall" and recommendations by "should". This standard highlights the need for complete and precise requirements, derived from the plant safety goals, as a pre-requisite for generating the comprehensive requirements for the total I&C system architecture, and hence for the individual I&C systems important to safety. This standard introduces the concept of a safety life cycle for the total I&C system architecture, and a safety life cycle for the individual systems. The life cycles illustrated in, and followed by, this standard are not the only ones possible; other life cycles may be followed, provided that the objectives stated in this standard are satisfied. 2 Application: new and pre-existing plants This standard applies to the I&C of new nuclear power plants as well as to I&C up-grading or back-fitting of existing plants. For existing plants, only a subset of requirements is applicable and this subset is identified at the beginning of any project. 3 Framework Figure 1 presents the overall framework of this standard, with its normative clauses: · clause 5 addresses the total architecture of the I&C systems important to safety: - defining requirements for the I&C functions, and associated systems and equipment (I&C FSE) derived from the safety analysis of the NPP, the categorisation of I&C functions, and the plant lay-out and operation context; - structuring the totality of the I&C architecture, dividing it into a number of systems and assigning the I&C functions to systems. Design criteria are identified, including those to give defence in depth and to minimise potential for common cause failure (CCF); - planning the total architecture of I&C systems. · clause 6 addresses the requirements for the individual I&C systems important to safety, particularly the requirements for computer-based systems; · clauses 7 and 8 address the overall integration, commissioning, operation and maintenance of the I&C systems; · annex A highlights the relations between IAEA and basic safety concepts that are used throughout this standard; · annex B provides information on the categorisation/classification principles; · annex C gives examples of I&C sensitivity to CCF; · annex D provides guidance to support comparison of this standard with parts 1, 2 and 4 of IEC 61508. This annex surveys the main requirements of IEC 61508 to verify that the issues relevant to safety are adequately addressed, considers the use of common terms and explains the reason for adopting different or complementary techniques or terms.

Nuclear power plant - Instrumentation and control for systems important to safety; General requirements for systems

NF C01-393/A1:2001 电工词汇.393章.核测量仪.物理现象和基本概念

Electrotechnical vocabulary - Chapter 393 : nuclear instrumentation : physical phenomena and basic concepts.

IEC TR 61838:2001 核电厂 对安全重要的仪器仪表和控制功能元件 分类用概率统计安全评估的使用

This Technical Report provides a survey of some of the methods by which probabilistic risk assessment results can be used to establish "risk based" classification criteria, so as to allow FSEs to be placed within the four categories established within IEC 61226. The application of risk based techniques, in conjunction with the consequence based classification approach given in IEC 61226, is currently decided by the utility and/or regulator within member Nations. In the absence of an internationally agreed approach, this should continue, but this Technical Report is intended to stimulate debate on this subject and encourage the convergence of views so that an IEC International Standard can be agreed. The safety principles and the usefulness of a risk-based approach to classification are discussed and a description of four different approaches is presented. In other respects, references are given in this report to IEC and IAEA documents which relate directly to the topic. This report also discusses the limitations associated with the use of probabilistic safety assessment (PSA) techniques. Guidance is given in annex A on modelling instrumentation and control functions for probabilistic risk assessment.

Nuclear power plants - Instrumentation and control functions important for safety - Use of probabilistic safety assessment for the classification

ASME BPVC CCB NC-2001 ASME锅炉和压力容器规程.标准件:CCB NC.核元件

ASME Boiler & Pressure Vessel Code - Code Cases: CCB NC; Nuclear Components

IEEE 420-2001 核电站用1E级控制板、操纵台及操纵仪表架的设计和鉴定

This standard applies to the design and qualification of Class 1E control boards, panels, and racks. It does not apply to individual components, modules, and external field-run cables except as they may affect the design and qualification of Class 1E control boards, panels, and racks.

Standard for the Design and Qualification of Class 1E Control Boards, Panels, and Racks Used in Nuclear Power Generating Stations

ANSI/IEEE 628:2001 核电站1E级电路用管道系统的设计、安装和合格鉴定的标准

Criteria for rhe minimum requirements in the selection, design, installation, and qualification of raceway systems for Class 1E circuits for nuclear power generating stations is provided.

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

ASTM E853-01(2008) E706(IA)轻水堆监测结果的分析和说明标准方法

The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to 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. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life. To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the materials test specimens were exposed. The resultant information will then become part of a data base applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole. To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67). The objectives and requirements of a reactor vessel''s support structure''s surveillance program are much less stringent, and at present, are limited to physics-dosimetry measurements through ex-vessel cavity monitoring coupled with the use of available test reactor metallurgical data to determine the condition of any support structure steels that might be subject to neutron induced property changes (1, 29, 44-58, 65-70). 1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70). 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. Note 18212;(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity). 1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E 560, Practice E 1006, Guide E 900

Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)