27.120.10 反应堆工程 标准查询与下载

DB34/T 3574.1-2019 聚变装置铁磁性系统结构设计准则 第1部分：总体设计

本部分规定了聚变装置铁磁性系统总体设计中的术语和定义、设计要求、RAMI要求、安全要求。 本部分适用于聚变装置铁磁性系统结构设计的所有阶段，是包层系统、偏滤器系统和管林系统等在 设计、制造检测及辐射屏蔽的总体要求。

Structural Design Criteria for Ferromagnetic Systems of Fusion Devices Part 1: Overall Design

KS C IEC 60744-2019 核电厂安全逻辑组件 特性及试验方法

Safety logic assemblies of nuclear power plants－Characteristics and test methods

ASTM C781-19 气冷核反应堆部件用石墨材料试验的标准实施规程

Standard Practice for Testing Graphite Materials for Gas-Cooled Nuclear Reactor Components

ASTM E2215-19 从轻水慢化核动力反应堆容器中评估监视胶囊的标准实施规程

1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules. 1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life. 1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel. 1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any lightwater moderated nuclear reactor.2 1.5 Modifications to the standard test program and supplemental tests are described in Guide E636. 1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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

ISO 10979:2019 识别核动力堆燃料组件

This document specifies requirements for the unique identification of fuel assemblies utilized in nuclear power plants. It was developed primarily for commercial light-water reactor fuel, but can be used for any reactor fuel contained in discrete fuel assemblies that can be identified with an identification code as specified by this document. This document defines the characters and proposed sequence to be used in assigning identification to the fuel assemblies. The identification is intended to be borne by the fuel assembly throughout its lifetime. This document aims at providing an organizing principle for fuel assembly identification systems in order to guarantee unequivocal identification at any time and any place in the world (see also IAEA Safety Guide No. GS-G-3.5). Considering that existing standards for fuel assembly identification (such as ANSI/ANS-57.8-1995, DIN 25433, IAEA Safety Guide No. GS-G-3.5) ensure unequivocal identification in their respective fields of application, this document allows without restriction the further application of these standards. Moreover, it is intended that this document be used as a guideline for new definitions of identification systems.

Identification of fuel assemblies for nuclear power reactors

NB/T 20522-2018 核电厂堆芯中子通量测量指套管涡流检测

Nuclear power plant core neutron flux measurement refers to casing eddy current testing

NB/T 20526-2018 压水堆蒸汽发生器出厂水压试验要求

Requirements for factory hydraulic pressure test of pressurized water reactor steam generators

NB/T 20515-2018 压水堆核电厂乏燃料组件湿法贮存临界安全分析准则

Criticality safety analysis criteria for wet storage of spent fuel assemblies in pressurized water reactor nuclear power plants

ASTM C781-18 气冷核反应堆部件用石墨材料试验的标准实施规程

Standard Practice for Testing Graphite Materials for Gas-Cooled Nuclear Reactor Components

DB34/T 3186.1-2018 聚变堆遥操作维护系统设计准则 第1部分：郑体维护要求及分析

DB34/T 3186 的本部分规定了聚变堆遥操作维护系统的术语和定义、聚变堆遥操作维护系统组成、整体维护要求、RAMI 要求和安全要求。 本部分适用于聚变堆遥操作维护系统设计的所有阶段，是遥操作维护子系统设计的重要依据。

Fusion Reactor Teleoperation and Maintenance System Design Criteria Part 1: Systematic Maintenance Requirements and Analysis

DB34/T 3186.7-2018 聚变堆遥操作维护系统设计准则 第7部分：多功能机械臂系统

DB34/T 3186 的本部分规定了聚变堆多功能机械臂系统在设计阶段所需考虑的术语和定义、系统功能及基本组成、系统设计、工程设计、RAMI 要求和安全要求。 本部分适用于聚变堆遥操作维护系统的设计。

Fusion Reactor Teleoperation and Maintenance System Design Guidelines Part 7: Multifunctional Manipulator System

DB34/T 3186.5-2018 聚变堆遥操作维护系统设计准则 第5部分：包层遥操作维护系统

DB34/T 3186 的本部分规定了聚变堆包层遥操作维护系统的术语和定义、系统功能及基本组成、系统设计、工程设计、RAMI 要求和安全要求。 本部分适用于聚变堆包层遥操作维护系统设计的所有阶段，是该系统设计的重要依据。

Fusion Reactor Teleoperation Maintenance System Design Guidelines Part 5: Cladding Teleoperation Maintenance System

DB34/T 3186.6-2018 聚变堆遥操作维护系统设计准则 第6部分：第一壁部件远程监控及测量系统

DB34/T 3186 的本部分规定了聚变堆第一壁部件远程监控及测量系统的术语和定义、系统功能及基本组成、系统设计、工程设计、RAMI 要求和安全要求。 本部分适用于聚变堆第一壁部件远程监控及测量系统设计的所有阶段，是该系统设计的重要依据。

Fusion Reactor Teleoperation and Maintenance System Design Guidelines Part 6: Remote Monitoring and Measurement System for First Wall Components

DB34/T 3186.3-2018 聚变堆遥操作维护系统设计准则 第3部分：转运车系统

DB34/T 3186 的本部分规定了聚变堆转运车系统的相关术语和定义、系统功能及基本组成、系统设计、工程设计、RAMI 要求和安全要求。 本部分适用于聚变堆转运车系统设计的所有阶段。

Fusion Reactor Teleoperation and Maintenance System Design Guidelines Part 3: Transfer Vehicle System

DB34/T 3186.2-2018 聚变堆遥操作维护系统设计准则 第2部分：遥操作协同控制系统

DB34/T 3186 的本部分规定了聚变堆遥操作协同控制系统的术语和定义、系统功能及基本组成、系统要求、系统设计、RAMI 要求和安全要求。 本部分适用于聚变堆遥操作协同系统设计的所有阶段。

Fusion Reactor Teleoperation and Maintenance System Design Guidelines Part 2: Teleoperation Cooperative Control System

DB34/T 3186.4-2018 聚变堆遥操作维护系统设计准则 第4部分：偏滤器遥操作维护系统

DB34/T 3186 的本部分规定了聚变堆偏滤器遥操作维护系统的术语和定义、系统功能及基本组成、系统要求、系统设计、RAMI 要求和安全要求。 本部分适用于聚变堆偏滤器遥操作维护系统设计阶段。

Fusion Reactor Teleoperation Maintenance System Design Guidelines Part 4: Divertor Teleoperation Maintenance System

ASTM E2215-18 从轻水慢化核动力反应堆容器中评估监视胶囊的标准实施规程

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

ISO 18077:2018 压水反应堆的重新加载启动物理试验

Reload startup physics tests for pressurized water reactors

NB/T 20507-2018 压水堆核电厂启动给水系统设计准则

Design criteria for start-up water supply system of pressurized water reactor nuclear power plant

ISO 18075:2018 用于动力反应堆分析的稳态中子学方法

This document provides guidance for performing and validating the sequence of steady-state calculations leading to prediction, in all types of operating UO2-fuel commercial nuclear reactors, of: — reaction-rate spatial distributions; — reactivity; — change of nuclide compositions with time. The document provides: a) guidance for the selection of computational methods; b) criteria for verification and validation of calculation methods used by reactor core analysts; c) criteria for evaluation of accuracy and range of applicability of data and methods; d) requirements for documentation of the preceding.

Steady-state neutronics methods for power-reactor analysis