F60 核反应堆综合 标准查询与下载

NB/T 20193-2012 核电厂常规岛汽水管道设计技术规范

Technical specification for design of steam and water piping in conventional island of nuclear power plant

NB/T 20191-2012 压水堆核电厂结构设计中在役检查的可达性准则

Accessibility Criteria for In-service Inspection in Structural Design of PWR Nuclear Power Plant

ANSI/ANS 6.1.2-2013 核电站核辐射防护和屏蔽计算用的中子和γ射线截面

Neutron and Gamma-Ray Cross Sections for Nuclear Radiation Protection and Shielding Calculations for Nuclear Power Plants

ASTM E1035-13 测定核反应堆容器支座结构中子辐照的标准实施规程

3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 2, 3, 4, 5).6 The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports. 3.2 It is known that high energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 ?? 10-4, used in this practice for neutron exposures. (See 1.2). 1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for: 1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities. 1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures. 1.2 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E gt; 1 MeV) that exceeds 18201;??8201;1017 neutrons/cm 2 or 3.08201;??8201;10???4 dpa.2 (See Terminology E170.) 1.3 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2. 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 and health prac......

Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures

IEEE 420-2013 核电站中使用的1E类别控制板,面板和机框限制条件和设计用IEEE标准

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

ASTM D6071-13 用石墨反应堆原子吸收光谱法测定高纯度水中低水平钠的标准试验方法

5.1 Small quantities of sodium, 1 to 10 μg/L, can be significant in high pressure boiler systems and in nuclear power systems. Steam condensate from such systems must have less than 10 μg/L. In addition, condensate polishing system effluents should have less than 1 μg/L. Graphite furnace atomic absorption spectroscopy (GFAAS) represents technique for determining low concentrations of sodium. 1.1 This test method covers the determination of trace sodium in high purity water. The method range of concentration is 1 to 40 μg/L sodium. The analyst may extend the range once its applicability has been ascertained. Note 1—It is necessary to perform replicate analysis and take an average to accurately determine values at the lower end of the stated range. 1.2 This test method is a graphite furnace atomic absorption spectrophotometric method for the determination of trace sodium impurities in ultra high purity water. 1.3 This test method has been used successfully with a high purity water matrix.2 It is the responsibility of the analyst to determine the suitability of the test method for other matrices. 1.4 The values stated in SI units are to be regarded as standard. 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 Test Method for Low Level Sodium in High Purity Water by Graphite Furnace Atomic Absorption Spectroscopy

ASTM B351/B351M-13 核设备用热轧和冷加工锆和锆合金棒材、杆材和线材的标准规格

1.1 This specification covers four grades of wrought zirconium and zirconium alloy bars, rod, and wire as follows: 1.1.1 R60001—Unalloyed grade, 1.1.2 R60802—Zirconium-Tin alloy (Zircaloy 2), 1.1.3 R60804—Zirconium-Tin alloy (Zircaloy 4), and 1.1.4 R60901—Zirconium-Niobium alloy. 1.2 Unless a single unit is used, for example corrosion mass gain in mg/dm2, the values stated in either inch-pound or SI units are to be regarded separately as standard. The values stated in each system are not exact equivalents; therefore each system must be used independently of the other. SI values cannot be mixed with inch-pound values. 1.3 The following precautionary caveat pertains only to the test method portions of this specification. 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 Specification for Hot-Rolled and Cold-Finished Zirconium and Zirconium Alloy Bars, Rod, and Wire for Nuclear Application

ASTM E853-13 轻水反应堆监测结果的分析和说明的标准实施规程

3.1 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. 3.1.1 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. 3.1.2 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-

Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results

IEEE 317-2013 核能发电站外壳结构中电气穿入组件的IEEE标准

IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations

ASTM E1006-13 试验反应堆试验物理学计量测定结果的分析和说明的标准实施规程

3.1 The mechanical properties of steels and other metals are altered by exposure to neutron radiation. These property changes are assumed to be a function of chemical composition, metallurgical condition, temperature, fluence (perhaps also fluence rate), and neutron spectrum. The influence of these variables is not completely understood. The functional dependency between property changes and neutron radiation is summarized in the form of damage exposure parameters that are weighted integrals over the neutron fluence spectrum. 3.2 The evaluation of neutron radiation effects on pressure vessel steels and the determination of safety limits require the knowlege of uncertainties in the prediction of radiation exposure parameters (for example, dpa (Practice E693), neutron fluence greater than 1.0 MeV, neutron fluence greater than 0.1 MeV, thermal neutron fluence, etc.). This practice describes recommended procedures and data for determining these exposure parameters (and the associated uncertainties) for test reactor experiments. 3.3 The nuclear industry draws much of its information from databases that come from test reactor experiments. Therefore, it is essential that reliable databases are obtained from test reactors to assess safety issues in Light Water Reactor (LWR) nuclear power plants. 1.1 This practice covers the methodology summarized in Annex A1 to be used in the analysis and interpretation of physics-dosimetry results from test reactors. 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods. 1.3 Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. 1.4 This practice is directed towards the development and application of physics-dosimetry-metallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practices E185, E853, and E1035, Guides E900, E2005, E2006 and Test Method E646. 1.5 This standard may involve hazardous mate......

Standard Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments

ASTM E944-13 反应堆监测时中子光谱调节法应用的标准指南, E706 (IIA)

3.1 Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry measurements (see Test Method E1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any adjustment procedure can be only as reliable as are the input data. 3.2 Input Data and Definitions : 3.2.1 The symbols introduced in this section will be used throughout the guide. 3.2.2 Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols: These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1). 3.2.3 The neutron spectrum (see Terminology E170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of neutron energy E , is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or Monte Carlo, see Guide E482). The results of the calculation are customarily given in the form of multigroup fluences or fluence rates. where: Ej and Ej+1 are the lower and upper bounds for the j-th energy group, respectively, and k is the total number of groups. 3.2.4 The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section for the i-th reaction as a function of energy E will be denoted by the following:

Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E???706 (IIA)

KS C IEC 60231A:2012 核反应堆仪器通用原则.补充版1

이 표준은 원자력 발전소의 원자로 계측에 관한 일반원칙, 중자자 속 측정, 온도 측정, 냉

General principal of nuclear reactor instrumentation－First supplement to publication 60231(1967)

KS C IEC 61504:2012 核电站.安全的重要测量及控制系统.发电所放射线的监视

이 표준에서는 컴퓨터 기반 방사선 감시 계통(RMS)의 설계원칙 및 성능기준에 관한 지침을

Nuclear power plants－Instrumentation and control systems important to safety－Plant-wide radiation monitoring

KS C IEC 60231:2012 核反应堆仪器通用原理

1.2.1 이 권고사항은 원자로 계측에 대한 지침을 제공하고 좋은 관례로서의 기준을 제시한

General principles of nuclear reactor instrumentation

KS C IEC 60780:2012 核电厂.安全系统的电气设备.鉴定

이 표준은 원자력발전소에서 사용되는 안전계통 전기기기에 적용되며, 고장 시 안전계통의 성능

Nuclear power plants-Electrical equipment of the safety system-Qualification

KS C IEC 60772:2012 核电站外壳结构的电气穿入件

이 표준은 원자로 격납구조(이하 “격납구조”라 한다.) 내 케이블 관통부에 적용한다. 이

Electrical penetration assemblies in containment structures for nuclear power generating stations

KS C IEC 61771:2012 核电站.主要控制间.设计的鉴定和批准

이 표준은 원자력발전소의 제어실 계통 설계에 대한 확인 및 검증 절차를 규정하고, 기능 할

Nuclear power plants－Main control-room－Verification and validation of design

KS C IEC 61226:2012 核电厂.安全重要仪表和控制系统.仪表和控制功能分级

이 표준은 원자력발전소의 정보 및 명령 기능과 그 기능을 제공하는 I&C 계통 및

Nuclear power plants-Instrumentation and control important to safety-Classification of instrumentation and control functions

KS C IEC 60964:2012 核发电厂控制室的设计

이 표준은 원자력발전소 주 제어실의 인간-기계 연계와 관련된 요건을 제시한다. 이 표준은

Nuclear power plants－Control rooms－Design

KS C IEC 61227:2012 核电厂.控制室.操作员控制器

이 표준은 원자력발전소 제어실 설계에 적용되는 KS C IEC 60964를 보완하며, 개별

Nuclear power plants－Control rooms－Operator controls