GB/T 16702.1-2025Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants — Part 1: General principle (English PDF)

Foreword

This document is in accordance with the provisions of GB/T 1.1-2020 "Guidelines for standardization work Part 1: Structure and drafting rules for standardization documents" Drafting.

This document is part 1 of GB/T 16702 "Design Specifications for Mechanical Equipment of Nuclear Island in Pressurized Water Reactor Nuclear Power Plant".

The following parts are published.

— Part 1: General;

— Part 2: Class 1 equipment;

— Part 3: Level 2 equipment;

— Part 4: Level 3 equipment;

— Part 5: Small equipment;

— Part 6: Reactor internals;

— Part 7: Equipment support;

— Part 8: Low-pressure or atmospheric pressure storage tanks.

This document partially replaces GB/T 16702-2019 "Design Specification for Mechanical Equipment of Nuclear Island of Pressurized Water Reactor Nuclear Power Plant" and GB/T 16702- Compared with 2019, in addition to structural adjustments and editorial changes, the main technical changes are as follows.

— Changed the scope of application of this document (see Chapter 1, Chapter 1 of the 2019 edition);

— Added normative references to GB/T 4960.2, HAD003/08, HAF003, HAF601, and HAF603 (see Chapter 3, 6.8.1, Chapter 7, Appendix N);

— Deleted the relevant provisions of "Handling of Non-Conformities" (see 4.2 of the 2019 edition);

— Deleted the relevant provisions of "Parts List" and "Deviation Report" (see 4.3.2.2, 4.3.6.2 of the 2019 edition);

— Added relevant requirements for "small equipment" (see 5.2.1.2, Table 1);

— Changed the relevant requirements for "weld classification" (see 5.2.4, 4.4.2.5 of the 2019 edition);

— Added the main content of "Equipment Design Specification" (see 6.1);

— Deleted "Level 1 equipment", "Level 2 equipment", "Level 3 equipment", "small equipment", "core components", "equipment support", "low pressure or normal pressure storage "Cans" (see Chapters 5 to 11 of the 2019 edition);

— Changed the material grades and their corresponding material standard numbers, performance data, allowable stress values, fatigue curve data and fatigue curves (see

Appendix A , Appendix A of the 2019 edition);

— Added Table E.1, Table E.3, definitions of gasket parameters, forces required to ensure initial gasket sealing and forces required to ensure sealing of connection structures The force calculation formula is as follows (see Appendix E);

— Added the scope of reinforcement for takeover (see Appendix G);

— Changed the requirements on chemical elements and mechanical properties, deleted the use of stock materials, small batch material purchases and residual elements of materials Related requirements. Added the integrated lower head of the reactor pressure vessel (see Appendix M, Appendix M of the 2019 edition);

— Changed the requirements for heat treatment and welding process assessment, and added the measurement method and acceptance requirements for ferrite content (see Appendix N,

Appendix N of the 2019 edition);

— Added relevant provisions for "integral flat head with large opening" (see Appendix U).

Please note that some of the contents of this document may involve patents. The issuing organization of this document does not assume the responsibility for identifying patents.

This document was proposed and coordinated by the National Nuclear Energy Standardization Technical Committee (SAC/TC58).

This document was drafted by: China Nuclear Power Research and Design Institute, China Nuclear Power Engineering Co., Ltd., Shanghai Nuclear Engineering Research and Design Institute Co., Ltd.

Co., Ltd., China General Nuclear Power Engineering Co., Ltd., Nuclear and Radiation Safety Center of the Ministry of Ecology and Environment, China Machinery Productivity Promotion Center, Nuclear Industry Standardization Research Institute Research Institute.

The main drafters of this document are: Huang Zongren, Li Changxiang, Fu Xiaolong, Tang Chuanbao, Hu Yu, Jing Yi, Tang Liang, Zuo Shuchun, Tang Yanjie, Yang Licai, Li Zheng, He Yinbiao, Zhang Yaochun, Deng Xiaoyun, Yin Qiwei, Wang Qingtian, Chen Hao, Zhang Kefeng, Yang Zhihai, Zhang Kai, Liu Hongbin, Wu Bingyang, Li Bingjin, Yu Xinyang, Tian Jun, Tang Huapeng, Wei Wei, Qiu Yang, Ma Shuli, Qiu Tian, Tan Bo, Li Donghui, Huang Min, Tang Chenhang, Su Yingbin, He Jinsong, Sun Yingxue, Yang Chuansheng, Shao Xuejiao, Du Juan, Lü Yongbo, Shi Kaikai, Zheng Bin, Xie Hai, Feng Zhipeng, Liu Wenjin, Zeng Zhongxiu, Li Hongying, Feng Lina, Wang Zhonghui, Lu Xiaohui, Wang Yanping, Qu Changming, Li Dongliang, Niu Yanying, Meng Xianggai, Zhang Jilai, Miao Ling, Qi Tao, Wang Zhenfeng, Song Yu, Zhang Junbao, Zhang Xing, Wang Yongdong, Gao Yongjian, Yin Haifeng, Shen Rui, You Lei, Gao Chen, Sheng Chaoyang, Zhou Zhou, Pan Jun, Wu Feifei, Li Cang.

The previous versions of this document and the documents it replaces are as follows.

— First published in 1996 as GB/T 16702-1996, first revised in 2019;

Introduction

GB/T 16702 "Design Specification for Mechanical Equipment of Nuclear Island of Pressurized Water Reactor Nuclear Power Plant" is a general technical specification for the design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant.

The standard is to implement the spirit of China's nuclear safety regulations, actively promote the unification of the standard technical route of mechanical equipment of the nuclear island of pressurized water reactor nuclear power plants, and promote the development of pressurized water reactors.

GB/T 16702 is an important part of the relevant standards for the independent design and localization of nuclear island mechanical equipment.

The basic and universal standards for design activities of mechanical equipment in the nuclear island of a pressurized water reactor nuclear power plant are planned to consist of eight parts.

— Part 1: General. The purpose is to specify the overall requirements and other related requirements for the design of mechanical equipment in the nuclear island of a pressurized water reactor nuclear power plant. Appendix for his partial use.

— Part 2: Class 1 equipment. The purpose is to specify the materials, design, manufacture, inspection, pressure test and overpressure of Class 1 pressure equipment.

Requirements that need to be followed in design such as protection.

— Part 3: Class 2 equipment. The purpose is to specify the materials, design, manufacture, inspection, pressure test and overpressure of Class 2 pressure equipment.

Requirements that need to be followed in design such as protection.

— Part 4: Class 3 equipment. The purpose is to specify the materials, design, manufacture, inspection, pressure test and overpressure of Class 3 pressure equipment.

Requirements that need to be followed in design such as protection.

— Part 5: Small equipment. The purpose is to specify the materials, design, manufacture, inspection, hydraulic test and pump of small pressure equipment.

Requirements that need to be followed in the design, such as identification and acceptance testing.

— Part 6: In-core components. The purpose is to specify the materials, design, manufacturing, inspection and other design requirements of in-core components. Require.

— Part 7: Equipment Support. The purpose is to specify the requirements to be followed in the design of mechanical equipment supports in the nuclear island of a pressurized water reactor nuclear power plant. Require.

— Part 8: Low-pressure or atmospheric pressure storage tanks. The purpose is to specify the materials, design, manufacture, inspection and hydraulic test of low-pressure or atmospheric pressure storage tanks.

Requirements that need to be followed in design, such as verification.

GB/T 16702 (all parts) together with NB/T.20001~NB/T.20009 series of standards constitute the pressurized water reactor nuclear power plant applicable to China.

The technical standard system for the design and manufacture of nuclear island mechanical equipment in power plants. This standard system is based on independent nuclear power engineering experience and incorporates nuclear island mechanical equipment The standard technical route unifies the research results, conforms to the requirements of China's nuclear power regulatory system and industrial foundation, and is the standard and guide for China's pressurized water reactor nuclear power plants.

It is an important basis for the design and manufacturing of nuclear island mechanical equipment and other related activities.

This document focuses on the differences between different design standard systems and China's engineering practices, adds some informative and normative appendices, and The material standard numbers and performance data were updated, thus improving the overall requirements for the design of mechanical equipment in the nuclear island of China's pressurized water reactor nuclear power plants.

Design specification for mechanical equipment of nuclear island of pressurized water reactor nuclear power plant Part 1: General

1 Scope

This document specifies the general requirements for the design of mechanical equipment in the nuclear island of a pressurized water reactor nuclear power plant.

This document is applicable to the design of nuclear island mechanical equipment (pressure-bearing equipment and its supports, and reactor internals) in pressurized water reactor nuclear power plants.

2 Normative references

GB/T 1954

GB/T 4960.2

GB/T 16702

GB/T 16702.2-2025

GB/T 16702.3-2025

GB/T 16702.4-2025

GB/T 16702.5-2025

GB/T 16702.6-2025

GB/T 16702.7-2025

GB/T 16702.8-2025

GB/T 25778