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GB/T 16702.4-2025: Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 4: Class 3 components
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Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 4: Class 3 components
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GB/T 16702.4-2025
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Basic data | Standard ID | GB/T 16702.4-2025 (GB/T16702.4-2025) | | Description (Translated English) | Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 4: Class 3 components | | Sector / Industry | National Standard (Recommended) | | Classification of Chinese Standard | F69 | | Classification of International Standard | 27.120.20 | | Word Count Estimation | 26,268 | | Date of Issue | 2025-02-28 | | Date of Implementation | 2025-02-28 | | Issuing agency(ies) | State Administration for Market Regulation, China National Standardization Administration | GB/T 16702.4-2025: Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 4: Class 3 components
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GB/T 16702.4-2025 English version. Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 4.Class 3 components
ICS 27.120.20
CCSF69
National Standard of the People's Republic of China
Partially replaces GB/T 16702-2019
Design specification for mechanical equipment of nuclear island of pressurized water reactor nuclear power plant
Part 4.Class 3 equipment
waterreactornuclearpowerplants-Part 4.Class3components
Released on 2025-02-28
2025-02-28 Implementation
State Administration for Market Regulation
The National Standardization Administration issued
Table of Contents
Preface III
Introduction IV
1 Scope 1
2 Normative references 1
3 Terms and Definitions 2
4 General 2
4.1 Documents to be prepared 2
4.2 Boundaries of equipment and piping components and accessories 2
4.3 Identification 6
5 Materials 6
5.1 Overview 6
5.2 Principles of material selection 6
5.3 Intergranular corrosion sensitivity 10
5.4 Cobalt content of austenitic and austenitic-ferritic stainless steels and nickel-chromium-iron alloys 11
5.5 Classification of pump internal components (structural categories) 12
6 Design12
6.1 General design principles 12
6.2 Vessel design rules 12
6.3 Pump design rules 14
6.4 General design rules for valves 14
6.5 Pipeline Design 18
7 Manufacturing and Inspection 19
8 Pressure test 19
9 Overpressure protection 19
Reference 20
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 4 of GB/T 16702 "Design Specification 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 replaces Chapter 7 Level 3 Equipment of GB/T 16702-2019 "Design Specifications for Mechanical Equipment of Nuclear Island of Pressurized Water Reactor Nuclear Power Plant".
Compared with Chapter 7 in GB/T 16702-2019, in addition to structural adjustments and editorial changes, the main technical changes are as follows.
--- Increased the scope of applicable jurisdiction and detailed provisions on the boundaries of level 3 equipment components and accessories (see 4.2);
--- Changed some of the adopted material standards, adopted the material standard unification and material branding carried out in the unification of the two technical routes
The results of the unification of the number (see 5.2, 7.2.2 of the.2019 edition);
--- Added cobalt content requirements based on usage scenarios (see 5.4.3);
--- Added welding joint coefficients corresponding to different welding methods and non-destructive testing ratios (see 6.2.2).
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 is proposed and coordinated by the National Nuclear Energy Standardization Technical Committee (SAC/TC58).
This document was drafted by. China Nuclear Power Engineering Co., Ltd., Shanghai Nuclear Engineering Research and Design Institute Co., Ltd., China Nuclear Power Research Institute
Design Institute, 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. Zuo Shuchun, Zhu Jingmei, Jin Wenjuan, Meng Xianggai, Miao Ling, Zhou Quan, Zhang Yaochun, Yang Chuansheng, Dong An, Zhang Feng,
Liu Jiayi, Gong Zhao, Sun Jiali, Gao Wenjia, Wang Chao, Fan Zhang, Sheng Feng, Wang Zhenfeng, You Yan, Gao Yongjian, Gu Chunhui, Ni Yiyu, Song Yu, Tian Yajing,
Fu Xiaolong, Lin Yishan, Feng Juanjuan, Sun Yiyun, Sheng Chaoyang, Gao Chen, Su Xihui, Li Yingke, Wu Feifei, Li Kang.
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 my country'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 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 my country.
The technical standard system for the design and manufacture of nuclear island mechanical equipment in power plants is based on independent nuclear power engineering experience and incorporates the experience of nuclear island mechanical design and manufacture.
The research results of the unified standard technical route are in line with the requirements of my country's nuclear power regulatory system and industrial foundation, and are the norms and guidance for my country's pressurized water reactor nuclear power
It is an important basis for the design and manufacturing of nuclear island mechanical equipment and other related activities.
This document focuses on the design principles of Class 3 pressure equipment, adds design requirements for special components and structures, and updates material standards.
The standard improves the requirements for materials, design, manufacturing, inspection, pressure testing and overpressure protection of Class 3 pressure equipment.
This document is used in conjunction with GB/T 16702.1-2025 and GB/T 16702.3-2025.
Design specification for mechanical equipment of nuclear island of pressurized water reactor nuclear power plant
Part 4.Class 3 equipment
1 Scope
This document specifies the materials, design, manufacturing, inspection, testing and overpressure protection of Class 3 equipment in the nuclear island mechanical equipment of a pressurized water reactor nuclear power plant.
Require.
This document applies to Level 3 pressure equipment and its components of pressurized water reactor nuclear power plants specified in GB/T 16702.1-2025.
2 Normative references
The contents of the following documents constitute essential clauses of this document through normative references in this document.
For referenced documents without a date, only the version corresponding to that date applies to this document; for referenced documents without a date, the latest version (including all amendments) applies to
This document.
GB/T 16702.1-2025 Specification for design of mechanical equipment for pressurized water reactor nuclear island Part 1.General
GB/T 16702.2-2025 Specification for design of mechanical equipment for pressurized water reactor nuclear island Part 2.Level 1 equipment
GB/T 16702.3-2025 Specification for design of mechanical equipment for pressurized water reactor nuclear island Part 3.Level 2 equipment
GB/T 16702.7 Specification for design of mechanical equipment for pressurized water reactor nuclear island Part 7.Equipment support
NB/T.20001-2023 Specification for the manufacture of mechanical equipment for nuclear island of pressurized water reactor nuclear power plant
NB/T.20002 (all parts) Specification for welding of mechanical equipment in nuclear island of pressurized water reactor nuclear power plant
NB/T.20005.1 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 1.Forgings and rolled products for level 1, 2 and 3 equipment
NB/T.20005.2 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 2.Tube sheet forgings for 2nd and 3rd stage heat exchangers
NB/T.20005.5 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 5.Grade 1, 2 and 3 pressure-bearing castings
NB/T.20005.7 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 7.Steel plates for level 1, 2 and 3 equipment
NB/T.20005.9 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 9.Seamless steel tubes for level 2 and 3 equipment
NB/T.20005.10 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 10.Grade 2 and 3 steel pipes welded with filler metal
NB/T.20005.13 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 13.Seamless steel for heat transfer tubes in 2nd and 3rd stage heat exchangers
Steel pipe
NB/T.20005.14 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 14.Grade 2 and 3 butt-weld seamless pipe fittings
NB/T.20005.15 Carbon steel and low alloy steel for pressurized water reactor nuclear power plants Part 15.Grade 2 and 3 pipe fittings welded with filler metal
NB/T.20007.1 Stainless steel for pressurized water reactor nuclear power plants Part 1.Austenitic stainless steel forgings for level 1, 2 and 3 equipment
NB/T.20007.2 Stainless steel for pressurized water reactor nuclear power plants Part 2.Austenitic stainless steel forgings for tube sheets of 2nd and 3rd stage heat exchangers
NB/T.20007.5 Stainless steel for pressurized water reactor nuclear power plants Part 5.Austenitic stainless steel plates for level 1, 2 and 3 equipment
NB/T.20007.8 Stainless steel for pressurized water reactor nuclear power plants Part 8.Austenitic stainless steel seamless steel pipes for level 1, 2 and 3 equipment
NB/T.20007.9 Stainless steel for pressurized water reactor nuclear power plants Part 9.Grade 1, 2, 3 austenitic stainless steel butt-welding seamless pipe fittings
NB/T.20007.10 Stainless steel for pressurized water reactor nuclear power plants Part 10.Seamless austenitic stainless steel for 1st, 2nd and 3rd stage heat exchangers
Steel pipe
NB/T.20007.11 Stainless steel for pressurized water reactor nuclear power plants Part 11.Grade 1, 2 and 3 austenitic stainless steel welded with filler metal
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