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GB/T 16702.1-2025 English PDF

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GB/T 16702.1-2025: Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General principle
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GB/T 16702.1: Evolution and historical versions

Standard IDContents [version]USDSTEP2[PDF] delivered inStandard Title (Description)StatusPDF
GB/T 16702.1-2025English6989 Add to Cart 25 days [Need to translate] Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General principle Valid GB/T 16702.1-2025
GB/T 16702-2019EnglishRFQ ASK 3 days [Need to translate] Design code for mechanical components in nuclear island of pressurized water reactor nuclear power plants Obsolete GB/T 16702-2019
GB/T 16702-1996EnglishRFQ ASK 3 days [Need to translate] Design rules for mechanical components of PWR nuclear islands Obsolete GB/T 16702-1996

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Basic data

Standard ID GB/T 16702.1-2025 (GB/T16702.1-2025)
Description (Translated English) Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General principle
Sector / Industry National Standard (Recommended)
Classification of Chinese Standard F69
Classification of International Standard 27.120.20
Word Count Estimation 374,324
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 16702.1-2025: Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General principle


---This is a DRAFT version for illustration, not a final translation. Full copy of true-PDF in English version (including equations, symbols, images, flow-chart, tables, and figures etc.) will be manually/carefully translated upon your order.
GB/T 16702.1-2025 English version. Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1.General principle 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 1.General waterreactornuclearpowerplants-Part 1.Generalprinciple 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 V 1 Scope 1 2 Normative references 1 3 Terms and definitions 3 4 General requirements 3 5 Equipment and its levels that comply with this document 4 5.1 Equipment that complies with this document 4 5.2 Equipment classification 4 6 Files 5 6.1 Equipment Design Specifications 5 6.2 Equipment Specifications 6 6.3 General technical documents 6 6.4 Documents related to procurement and manufacturing 6 6.5 Documents related to welding 7 6.6 Documents related to inspection 7 6.7 Non-conformity Report 8 6.8 Quality Documentation and Final Completion Report 8 7 Quality Assurance 9 8 Use of each part of GB/T 16702 9 Appendix A (Normative) Properties of design materials 10 Appendix B (Normative) Experimental Stress Analysis 54 Appendix C (Normative) Determination of basic allowable stress limits 63 Appendix D (Normative) Design rules for equipment subject to external pressure 65 Appendix E (Normative) Design of Circular Flange Bolted Connections 81 Appendix F (Normative) Linear support design rules 105 Appendix G (Normative) Rules for reinforcement of openings in Class 1 vessels 142 Appendix H (Informative) Fatigue Analysis of Significant Geometric Discontinuities 147 Appendix I (Informative) Other Rules for Level 1 Pipeline Analysis 150 Appendix J (Informative) Rules related to Class D criteria 162 Appendix K (informative) Prevention of rapid fracture 170 Appendix L (Informative) Rules to be followed in determining the service factor 180 Appendix M (Normative) Supplementary requirements for materials 184 Appendix N (Normative) Acceptance of support welding filler materials and welding procedure assessment 202 Appendix O (Informative) Stress Analysis Methods 207 Appendix P (Informative) Dynamic Analysis Method 256 Appendix Q (Informative) Design and evaluation of rectangular and circular cross-section welded attachments for Class 1, 2 and 3 pipelines 318 Appendix R (Informative) Permissible installation deviations for post-completion review of piping systems 332 Appendix S (informative) Class 2 and 3 flexible hoses with metal braiding 339 Appendix T (Informative) Design of FF-class flanges for Class 2 and 3 components 342 Appendix U (Normative) Integral flat head with large opening 361 References 364

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 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 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 my country. 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 my country's nuclear power regulatory system and industrial foundation, and is the standard and guide for my country'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 my country'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 my country'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

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 1954 Method for measuring ferrite content in chromium-nickel austenitic stainless steel welds GB/T 4960.2 Nuclear Science and Technology Terminology Part 2.Fission Reactor GB/T 16702 (all parts) Specification for design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant GB/T 16702.2-2025 Specification for design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant Part 2.Level 1 equipment GB/T 16702.3-2025 Specification for design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant Part 3.Level 2 equipment GB/T 16702.4-2025 Specification for design of mechanical equipment for nuclear island of pressurized water reactor nuclear power plant Part 4.Level 3 equipment GB/T 16702.5-2025 Specification for design of mechanical equipment for nuclear island of pressurized water reactor nuclear power plant Part 5.Small equipment GB/T 16702.6-2025 Specification for design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant Part 6.Reactor internals GB/T 16702.7-2025 Specification for design of mechanical equipment for nuclear island of pressurized water reactor nuclear power plant Part 7.Equipment support GB/T 16702.8-2025 Specification for design of mechanical equipment of nuclear island of pressurized water reactor nuclear power plant Part 8.Low pressure or atmospheric pressure storage tanks GB/T 25778 Welding Material Purchasing Guide NB/T.20001-2023 Specification for the manufacture of mechanical equipment for nuclear island of pressurized water reactor nuclear power plant NB/T.20002.1 Specification for welding of mechanical equipment in nuclear island of pressurized water reactor nuclear power plant Part 1.General requirements NB/T.20002.2 Specification for welding of mechanical equipment in nuclear island of pressurized water reactor nuclear power plant Part 2.Acceptance of welding filler materials NB/T.20002.3 Specification for welding of mechanical equipment in nuclear island of pressurized water reactor nuclear power plant Part 3.Welding procedure qualification NB/T.20002.6-2021 Specification for welding of mechanical equipment in nuclear island of pressurized water reactor nuclear power plant Part 6.Product welding NB/T.20003.1 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 1.General requirements NB/T.20003.2 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 2.Ultrasonic testing NB/T.20003.3 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 3.Radiographic testing NB/T.20003.4 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 4.Penetrant testing NB/T.20003.5 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 5.Magnetic particle testing NB/T.20003.6 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 6.Eddy current testing NB/T.20003.7 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 7.Visual inspection NB/T.20003.8 Nondestructive testing of mechanical equipment in nuclear island of nuclear power plant Part 8.Leakage detection 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

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