Standard No.:	GB/T 16702.6-2025
English Name:	Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants―Part 6: Reactor vessel internals
Chinese Name:	压水堆核电厂核岛机械设备设计规范第6部分：堆内构件
Professional Classification:	GB    National Standard
Source Content Issued by:	SAMR; SAC
Issued on:	2025-02-28
Implemented on:	2025-2-28
Status:	valid
Superseding:	GB/T 16702-2019 Design code for mechanical components in nuclear island of pressurized water reactor nuclear power plants
Target Language:	English
File Format:	PDF
Word Count:	20000 words
Translation Price(USD):	600.0
Delivery:	via email in 1~5 business day

GB/T 16702.6-2025 Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants―Part 6: Reactor vessel internals English, Anglais, Englisch, Inglés, えいご This is a draft translation for reference among interesting stakeholders. The finalized translation (passing through draft translation, self-check, revision and verification) will be delivered upon being ordered. ICS 27.120.20 CCS F 69 National Standard of the People's Republic of China GB/T 16702.6-2025 Partially replaces GB/T 16702-2019 Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 6: Reactor vessel internals 压水堆核电厂核岛机械设备设计规范 第6部分：堆内构件 (English Translation) Issue date: 2025-02-28 Implementation date: 2025-02-28 Issued by the State Administration for Market Regulation the Standardization Administration of the People's Republic of China Contents Foreword Introduction 1 Scope 2 Normative references 3 Terms and definitions 4 General principles 5 Materials 6 Design 7 Manufacturing and inspection Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 6: Reactor vessel internals 1 Scope This document specifies the requirements for materials, design, manufacturing, inspection, and other aspects of reactor vessel internals among the mechanical components in the nuclear island of pressurized water reactor nuclear power plants. This document applies to the design of reactor vessel internals and their components and parts. 2 Normative references The following documents contain requirements which, through reference in this text, constitute provisions of this document. For dated references, only the edition cited applies. For undated references, the latest edition of the referenced document (including any amendments) applies. GB/T 16702.1-2025 Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General principles GB/T 16702.2-2025 Design specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 2: Class 1 equipment NB/T 20001-2023 Manufacturing specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants NB/T 20002.1 Welding specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 1: General requirements NB/T 20002.5-2013 Welding specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 5: Manufacturing shop qualification NB/T 20002.6-2021 Welding specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 6: Product welding NB/T 20002.7 Welding specification for mechanical components in nuclear island of pressurized water reactor nuclear power plants - Part 7: Wear-resistant surfacing welding NB/T 20003.1-2021 Non-destructive testing for mechanical components in nuclear island of nuclear power plants - Part 1: General requirements NB/T 20003.2-2021 Non-destructive testing for mechanical components in nuclear island of nuclear power plants - Part 2: Ultrasonic testing NB/T 20003.3 Non-destructive testing for mechanical components in nuclear island of nuclear power plants - Part 3: Radiographic testing NB/T 20003.4 Non-destructive testing for mechanical components in nuclear island of nuclear power plants - Part 4: Penetrant testing NB/T 20004-2014 Methods for physical and chemical inspection of materials for mechanical components in nuclear island of nuclear power plants 3 Terms and definitions The terms and definitions defined in GB/T 16702.1-2025 apply to this document. 4 General principles 4.1 Documents Documents shall be implemented in accordance with the requirements of Clause 6 in GB/T 16702.1-2025. 4.2 Scope and classification of reactor vessel internals 4.2.1 Scope of reactor vessel internals All components and parts within the reactor pressure vessel, excluding fuel assemblies and their associated components, control rod drive mechanisms, irradiation surveillance tubes, and core measurement elements, are classified as reactor vessel internals. The designer shall specify their material selection standards, manufacturing grades, and corresponding service limits based on the functions of the reactor vessel internals. 4.2.2 Classification of reactor vessel internals All components and parts of reactor vessel internals shall be divided into two categories: Core Support Structures (CS) and In-Vessel Structures (IS) in accordance with the requirements of 5.2.2.1 in GB/T 16702.1-2025. a) Core Support Structures refer to the structural components in the reactor that provide support and positioning constraints for the fuel assemblies forming the core. Structures that provide support and limit constraints for the core only after a hypothetical failure accident of the core support structure belong to the reactor vessel internals. b) All components of reactor vessel internals except Core Support Structures are classified as In-Vessel Structures. c) The connecting welds between In-Vessel Structures and Core Support Structures belong to Core Support Structures. d) Temporary fasteners that are in contact with or connected to reactor vessel internals and then removed do not belong to reactor vessel internals. Temporary fasteners include base plates for pre-service instruments, tie rods, struts and protective covers, and hoisting fasteners for centering, etc. 4.2.3 Boundary division The boundary between the Core Support Structure and the reactor pressure vessel shall be located on the surface of the Core Support Structure. The first connecting weld between the Core Support Structure and the reactor pressure vessel shall be considered part of the reactor pressure vessel, unless the distance between the weld and the pressure-bearing area of the reactor pressure vessel is greater than 2t (t is the nominal thickness of the pressure-bearing material). Unless otherwise specified in the specification, the mechanical connection structure used to connect the Core Support Structure and the reactor pressure vessel is within the scope of this document. 4.3 Identification 4.3.1 Purpose of equipment identification The component identification system is an identification (equipment traceability) system that establishes a clear connection between the equipment components or welding joints and a set of related documents. 4.3.2 Adaptability of identification system to production management methods The component identification system shall match the type of management method adopted for the parts or welds: for single item control, each item shall use an identification number; for batch control, each batch shall adopt an identification number. 4.3.3 Identification methods The methods used to identify components include: a) Etching methods (including stamping, scribing, etc.); b) Temporary marking methods (using ink, paint, etc.);