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ASME BPVC XI 2 2023

$377.00

ASME BPVC.XI.2-2023 Section XI, Rules for Inservice Inspection of Nuclear Reactor Facility Components, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Reactor Facilities

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ASME 2023 151
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Provides requirements to maintain the nuclear power plant while in operation and to return the plant to service following plant outages. The rules require a mandatory program to evidence adequate safety and manage deterioration and aging effects. The rules also stipulate duties of the Authorized Nuclear Inservice Inspector to verify that the mandatory program has been completed, permitting the plant to return to service in a safe and expeditious manner. Application of this Section begins when the requirements of the construction code have been satisfied. DIVISION 2 This Division provides the requirements for the creation of the Reliability and Integrity Management (RIM) Program for advanced nuclear reactor designs. The RIM Program addresses the entire life cycle for all types of nuclear power plants, it requires a combination of monitoring, examination, tests, operation, and maintenance requirements that ensures each Structure, System, and Component (SSC) meets plant risk and reliability goals that are selected for the RIM Program.

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PDF Pages PDF Title
5 TABLE OF CONTENTS
10 List of Sections
11 FOREWORD
13 STATEMENT OF POLICY ON THE USE OF THE ASME SINGLE CERTIFICATION MARK AND CODE AUTHORIZATION IN ADVERTISING
STATEMENT OF POLICY ON THE USE OF ASME MARKING TO IDENTIFY MANUFACTURED ITEMS
14 Personnel
36 Correspondence With the Committee
38 PREFACE TO SECTION XI
INTRODUCTION
GENERAL
39 ORGANIZATION OF SECTION XI
1 DIVISIONS
2 ORGANIZATION OF DIVISION 1
40 3 ORGANIZATION OF DIVISION 2
41 4 REFERENCES
43 SUMMARY OF CHANGES
44 Cross-Referencing in the ASME BPVC
45 ARTICLE RIM-1 SCOPE AND RESPONSIBILITY
RIM-1.1 Scope
RIM-1.2 Jurisdiction
RIM-1.3 Components Subject to the Requirements of This Division
RIM-1.4 Owner’s Responsibility
46 RIM-1.5 Standard Units
RIM-1.6 Inspection
RIM-1.6.1 Duties of the Inspector and Authorized Nuclear Inservice Inspector Supervisor
RIM-1.6.2 Qualification of Authorized Inspection Agencies, Inspectors, and Supervisors
RIM-1.6.3 Access for Inspector
RIM-1.7 Regulatory Review
RIM-1.8 Tolerances
47 RIM-1.9 Referenced Standards and Specifications
Tables
Table RIM-1.9-1 Referenced Standards and Specifications
48 ARTICLE RIM-2 RELIABILITY AND INTEGRITY MANAGEMENT (RIM) PROGRAM
RIM-2.1 RIM Program Overview
RIM-2.1.1 Basis, Objective, and Process
RIM-2.1.2 Responsibilities
RIM-2.2 RIM Program Scope and Definition
RIM-2.3 Degradation Mechanism Assessment (DMA)
49 RIM-2.4 Facility and SSC Reliability Target Allocation
RIM-2.4.1 Facility-Level Risk and Reliability Targets
RIM-2.4.2 SSC-Level Reliability Targets
RIM-2.4.3 Scope, Level of Detail, and Technical Adequacy of the PRA
RIM-2.5 Identification and Evaluation of RIM Strategies
RIM-2.5.1 Identification of RIM Strategies
50 RIM-2.5.2 Evaluation of RIM Strategy Impacts on SSC Reliability
RIM-2.6 Evaluation of Uncertainties
RIM-2.7 RIM Program Implementation
RIM-2.7.1 RIM Program Documentation
RIM-2.7.2 Inservice Inspection Interval
51 RIM-2.7.3 Preservice Inspection
RIM-2.7.4 Design Requirements for RIM
RIM-2.7.5 Leak Detection System Requirements for RIM
RIM-2.7.6 MANDE Requirements for RIM
53 RIM-2.7.7 Examination Methods and Volumes
RIM-2.8 Performance Monitoring and RIM Program Updates
RIM-2.9 Examination Methods
RIM-2.9.1 Visual Examinations
54 RIM-2.9.2 Surface Examination
RIM-2.9.3 Volumetric Examination
55 RIM-2.9.4 Alternative Examinations
RIM-2.10 Additional Considerations for RIM Program Implementation
RIM-2.10.1 Consequence, External Event, and Shutdown Considerations
RIM-2.10.2 Principles of Risk-Informed Decision Making
57 ARTICLE RIM-3 ACCEPTANCE STANDARDS
RIM-3.1 Evaluation of Examination Results and Acceptance Standards
58 ARTICLE RIM-4 REPAIR/REPLACEMENT ACTIVITIES
RIM-4.1 Scope
RIM-4.2 Leakage Test Requirements After a Repair/Replacement Activity
RIM-4.2.1 Test Boundaries
RIM-4.2.2 Gas Leakage Test
RIM-4.2.3 Liquid Leakage Test
59 RIM-4.2.4 NDE in Lieu of Leakage Testing
RIM-4.2.5 Exemptions From Leakage Tests
RIM-4.3 Responsibilities
RIM-4.4 Corrective Action
RIM-4.5 Records
60 ARTICLE RIM-5 SYSTEM LEAKAGE MONITORING AND PERIODIC TESTS
RIM-5.1 Scope
RIM-5.2 Leakage Monitoring
RIM-5.2.1 General
RIM-5.2.2 Periodic Leakage Test
RIM-5.3 Corrective Action
RIM-5.4 Records
61 ARTICLE RIM-6 RECORDS AND REPORTS
RIM-6.1 Scope
RIM-6.2 Documentation Requirements
RIM-6.2.1 Owner’s Responsibilities
RIM-6.2.2 Owner’s Activity Report, Form OAR-1
RIM-6.2.3 Contracted Repair/Replacement Organization Responsibilities
RIM-6.2.4 Owners’ Repair/Replacement Certification Record NIS-2 Responsibilities
RIM-6.3 Record Retention
RIM-6.3.1 Maintenance of Records
RIM-6.3.2 Reproduction, Digitization, and Microfilming
RIM-6.3.3 Construction Records
RIM-6.3.4 RIM Program Records
62 RIM-6.3.5 Repair/Replacement Activity Records
63 ARTICLE RIM-7 GLOSSARY
RIM-7.1 Terms and Definitions
65 RIM-7.2 Acronyms
66 MANDATORY APPENDIX I RIM DECISION FLOWCHARTS FOR USE WITH THE RIM PROGRAM
ARTICLE I-1 FLOWCHARTS
I-1.1 General
67 Figures
Figure I-1.1-1 Inputs to the RIMEP for NPP Owner’s RIM Program Development
68 Figure I-1.1-2 RIM Program Development and Integration
69 Figure I-1.1-3 Process for Identifying the SSCs to Be in MANDE Program
70 Figure I-1.1-4 Selection of Strategies for SSCs to Meet Reliability Targets
71 Figure I-1.1-5 Upper Half Shows Input to MANDEEP for Developing MANDE Specification and Lower Half Shows Process for Evaluating if Section XI, Division 1 Requirements Meet MANDE Specifications
72 Figure I-1.1-6 Select, Develop, and Validate Performance Demonstration Approach to Meet SSC Reliability Target
73 MANDATORY APPENDIX II DERIVATION OF COMPONENT RELIABILITY TARGETS FROM FACILITY SAFETY REQUIREMENTS
ARTICLE II-1 GENERAL REQUIREMENTS
II-1.1 SCOPE
II-1.2 ADEQUACY OF THE PRA
II-1.3 PROCEDURE OVERVIEW
74 ARTICLE II-2 DERIVATION OF RELIABILITY TARGETS
II-2.1 Facility-Level Safety Requirements
II-2.2 Allocation of Reliability Targets
II-2.3 Identification of Component Groups
II-2.4 Trial Assignment of Reliability Targets
II-2.5 Evaluation of Impacts of Reliability Targets on Facility-Level Risk
II-2.6 Determination of Reliability Targets
75 MANDATORY APPENDIX III OWNER’S RECORD AND REPORT FOR RIM PROGRAM ACTIVITIES
ARTICLE III-1 GUIDES TO COMPLETING FORMS
III-1.1 Form OAR-1
III-1.2 Form NIS-2
76 Table III-1.1-1 Guide for Completing Form OAR-1
77 MANDATORY APPENDIX IV MONITORING AND NDE QUALIFICATION
ARTICLE IV-1 INTRODUCTION
IV-1.1 Scope
IV-1.2 Methods
IV-1.3 Owner’s Monitoring and NDE Expert Panel (MANDEEP)
IV-1.3.1 General Responsibilities
IV-1.3.2 MANDEEP-Specific Responsibilities
78 IV-1.3.3 MANDEEP Qualifications
79 ARTICLE IV-2 PERSONNEL QUALIFICATION
IV-2.1 Basic Personnel Qualification
IV-2.2 Method-Specific or Technique-Specific Personnel Qualifications
IV-2.2.1 Data Acquisition Personnel
IV-2.2.2 Data Evaluation Personnel
80 ARTICLE IV-3 MANDE METHODS AND TECHNIQUES RELIABILITY-BASED QUALIFICATION
IV-3.1 General
IV-3.2 Determination of the Qualification Requirements
IV-3.3 Qualification Process
IV-3.3.1 General
IV-3.3.2 SSC MANDE Specifications (Figure I-1.1-5)
IV-3.3.3 MANDE Technical Justification (Figure I-1.1-6)
IV-3.3.4 Levels of Rigor (Figure I-1.1-6)
IV-3.3.5 Qualification of NDE Methods and Techniques (Figure I-1.1-6)
81 IV-3.3.6 Monitoring Methods and Techniques (Figure I-1.1-6)
IV-3.3.7 Qualification Alternatives
82 ARTICLE IV-4 MANDE PERFORMANCE DEMONSTRATIONS (FIGURE I-1.1-6)
IV-4.1 General
IV-4.2 Personnel Performance Demonstration for Monitoring Methods
IV-4.3 NDE Personnel Performance Demonstration
IV-4.4 Procedure and Equipment Performance Demonstration
83 ARTICLE IV-5 RECORDS
IV-5.1 General
IV-5.2 Records for Methods and Technique Qualification
IV-5.3 Records for Personnel Performance Demonstrations
84 MANDATORY APPENDIX V CATALOG OF NDE REQUIREMENTS AND AREAS OF INTEREST
ARTICLE V-1 EXAMINATION CATEGORIES
V-1.1 Initial Consideration
Table V-1.1-1 Examination Category A, Pressure-Retaining Welds in Reactor Vessels
85 Table V-1.1-2 Examination Category B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels
Table V-1.1-3 Examination Category D, Full-Penetration Welded Nozzles in Vessels
86 Table V-1.1-4 Examination Category F, Pressure-Retaining Dissimilar Welds in Vessel Nozzles
87 Table V-1.1-5 Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter
88 Table V-1.1-6 Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) or Less in Diameter
89 Table V-1.1-7 Examination Category J, Pressure-Retaining Welds in Piping
90 Table V-1.1-8 Examination Category K, Welded Attachments for Vessels, Piping, Rotating Equipment, and Valves
Table V-1.1-9 Examination Category L-2, Pump Casings; Examination Category M-2, Valve Bodies
91 Table V-1.1-10 Examination Category N-1, Interior of Reactor Vessels; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; Examination Category N-3, Removable Core Support Structures
Table V-1.1-11 Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings
Table V-1.1-12 Examination Category P, All Pressure-Retaining Components
92 Table V-1.1-13 Examination Category F-A, Supports
93 MANDATORY APPENDIX VI RELIABILITY AND INTEGRITY MANAGEMENT EXPERT PANEL (RIMEP)
ARTICLE VI-1 OVERVIEW
VI-1.1 Responsibilities and Qualifications of RIMEP
94 MANDATORY APPENDIX VII SUPPLEMENTS FOR TYPES OF NUCLEAR REACTOR FACILITIES
ARTICLE VII-1 SUPPLEMENT FOR LIGHT WATER REACTOR–TYPE FACILITIES
VII-1.1 Scope
VII-1.2 RIM Program — Damage Degradation Assessment
VII-1.3 Acceptance Standards
VII-1.3.1 Evaluation of Examination Results
95 Table VII-1.2-1 Degradation Mechanism Attributes and Attribute Criteria (LWR)
102 VII-1.3.2 Supplemental Examinations
VII-1.3.3 Acceptance Standards
103 VII-1.3.4 Characterization
VII-1.3.5 Acceptability
VII-1.4 Acceptance Standards for Specific Examination Categories
VII-1.4.1 Acceptance Standards for Examination Categories A and B, Pressure-Retaining Welds in Reactor Vessel and Other Vessels
Table VII-1.3.3-1 Acceptance Standards
104 VII-1.4.2 Acceptance Standards for Examination Category D, Full-Penetration Welds of Nozzles in Vessels
105 VII-1.4.3 Acceptance Standards for Examination Category F, Pressure-Retaining Dissimilar Metal Welds in Vessel Nozzles and Category J, Pressure-Retaining Welds in Piping
106 VII-1.4.4 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter
VII-1.4.5 Acceptance Standards for Examination Category K, Welded Attachments for Vessels, Piping, Pumps, and Valves
107 VII-1.4.6 Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter, and Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) and Less in Diameter
VII-1.4.7 Acceptance Standards for Examination Categories L-2 and M-2, Equipment Casings and Valve Bodies
108 VII-1.4.8 Acceptance Standards for Examination Category N-1, Interior of Reactor Vessel; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; and Examination Category N-3, Removable Core Support Structures
VII-1.4.9 Acceptance Standards for Examination Category P, All Pressure-Retaining Components
VII-1.4.10 Acceptable Standards for Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings
109 VII-1.4.11 Acceptance Standards for Examination Category F-A, Component Supports
VII-1.5 Analytical Evaluation of Planar Flaws
110 VII-1.5.1 Acceptance Criteria for Ferritic Steel Components 4 in. (100 mm) and Greater in Thickness
VII-1.5.2 Acceptance Criteria for Ferritic Components Less Than 4 in. (100 mm) in Thickness
VII-1.5.3 Analytical Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping
111 VII-1.5.4 Evaluation Procedure and Acceptance Criteria for Head Penetration Nozzles of PWR Reactor Vessels
112 VII-1.6 Analytical Evaluation of Facility Operating Events
VII-1.6.1 Scope
VII-1.6.2 Unanticipated Operating Events
VII-1.6.3 Fracture Toughness Criteria for Protection Against Failure
VII-1.6.4 Operating Facility Fatigue Assessments
113 ARTICLE VII-2 SUPPLEMENT FOR LIQUID METAL REACTOR–TYPE FACILITIES
114 ARTICLE VII-3 SUPPLEMENT FOR HIGH-TEMPERATURE GAS REACTOR–TYPE FACILITIES
VII-3.1 Scope
VII-3.2 RIM Program — Damage Degradation Assessment
VII-3.3 Acceptance Standards
VII-3.3.1 Evaluation of Examination Results
115 Table VII-3.2-1 Degradation Mechanism Attributes and Attribute Criteria for High Temperature Gas Reactors
121 VII-3.3.2 Supplemental Examinations
VII-3.3.3 Acceptance Standards
VII-3.3.4 Characterization
VII-3.3.5 Acceptability
Table VII-3.3.3-1 Acceptance Standards
122 VII-3.4 Acceptance Standards for Specific Examination Categories
VII-3.4.1 Acceptance Standards for Examination Categories A and B, Pressure-Retaining Welds in Reactor Vessel and Other Vessels
VII-3.4.2 Acceptance Standards for Examination Category D, Full Penetration Welds of Nozzles in Vessels
123 VII-3.4.3 Acceptance Standards for Examination Category F, Pressure Dissimilar Metal Welds in Vessel Nozzles, and Examination Category J, Pressure-Retaining Welds in Piping
124 VII-3.4.4 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter
VII-3.4.5 Acceptance Standards for Examination Category K, Welded Attachments for Vessels, Piping, Pumps, and Valves
125 VII-3.4.6 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter, and Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) and Less in Diameter
VII-3.4.7 Acceptance Standards for Examination Categories L-2 and M-2, Rotating Equipment Casings and Valve Bodies
126 VII-3.4.8 Acceptance Standards for Examination Category N-1, Interior of Reactor Vessel; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; and Examination Category N-3, Removable Core Support Structures
VII-3.4.9 Acceptance Standards for Examination Category P, All Pressure-Retaining Components
VII-3.4.10 Acceptance Standards for Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings
127 VII-3.4.11 Acceptance Standards for Examination Category F-A, Component Supports
VII-3.5 Analytical Evaluation of Planar Flaws
128 VII-3.5.1 Acceptance Criteria for Ferritic Steel Components 4 in. (100 mm) and Greater in Thickness
129 VII-3.5.2 Acceptance Criteria for Ferritic Components Less Than 4 in. (100 mm) in Thickness
VII-3.5.3 Analytical Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping
130 VII-3.5.4 Evaluation Procedure and Acceptance Criteria for Head Penetration Nozzles of Reactor Vessels
VII-3.6 Analytical Evaluation of Facility Operating Events
VII-3.6.1 Scope
VII-3.6.2 Unanticipated Operating Events
VII-3.6.3 Fracture Toughness Criteria for Protection Against Failure
VII-3.6.4 Operating Facility Fatigue Assessments
132 ARTICLE VII-4 SUPPLEMENT FOR MOLTEN SALT REACTOR–TYPE FACILITIES
133 ARTICLE VII-5 SUPPLEMENT FOR GENERATION 2 LWR–TYPE FACILITIES
134 ARTICLE VII-6 SUPPLEMENT FOR FUSION MACHINE–TYPE FACILITIES
135 NONMANDATORY APPENDIX A ALTERNATE REQUIREMENTS FOR MONITORING AND NDE
ARTICLE A-1 GENERAL
A-1.1 Scope
A-1.2 Methods
A-1.3 Responsibilities
136 Figure A-1.2-1 Logic Flow Diagram of the Process for Determining Acceptability of Alternative Requirements
137 ARTICLE A-2 PROCEDURE FOR DETERMINING ACCEPTABILITY OF ALTERNATIVE MONITORING OR NDE
A-2.1 Overview
A-2.2 SSC Reliability Target
A-2.3 Degradation Mechanisms and Failure Modes
A-2.4 Approaches — Probabilistic and Deterministic
138 ARTICLE A-3 STAGE I EVALUATION
A-3.1 Introduction
A-3.2 Input Related to Safety Evaluation
A-3.3 Input Related to Structural Evaluation
A-3.4 Probabilistic Approach — Reliability Evaluation
A-3.4.1 Evaluation Procedure
A-3.4.2 Criteria
A-3.5 Deterministic Approach — Margin Assessment
A-3.5.1 Evaluation Procedure
A-3.5.2 Criteria
139 ARTICLE A-4 STAGE II EVALUATION
A-4.1 Introduction
A-4.2 Input Related to Safety Evaluation
A-4.3 Input Related to Structural Evaluation
A-4.4 Detectability
A-4.5 Criteria to Establish Additional Requirements
140 A-4.6 Probabilistic Approach
A-4.7 Deterministic Approach
141 ARTICLE A-5 PROCEDURE FOR STRUCTURAL RELIABILITY EVALUATION FOR PASSIVE COMPONENTS
A-5.1 General Requirements
A-5.1.1 Scope
A-5.1.2 Referenced Standards and Specifications
A-5.1.3 Application
A-5.2 Reliability Evaluation
A-5.2.1 General
A-5.2.2 Setting of Failure Scenario
Figure A-5.2.1-1 Reliability Evaluation Procedure
142 A-5.2.3 Modeling
A-5.2.4 Failure Probability Calculation
A-5.3 Setting of Failure Scenario
A-5.3.1 General
A-5.3.2 Setting of Limit State
A-5.3.3 Selection of Failure Modes
Figure A-5.3.1-1 Procedure for Setting Failure Scenarios
143 A-5.3.4 Setting of Failure Scenarios and Evaluation Portions
A-5.3.5 Setting of Reference Period
A-5.4 Modeling
A-5.4.1 General
A-5.4.2 Formulation of Scenario Developing Process
A-5.4.3 Setting of Limit-State Function
A-5.4.4 Setting of Random Variables
Figure A-5.4.1-1 Modeling Procedure
144 A-5.5 Reliability Calculation
145 ARTICLE A-6 RECORDS AND REPORTS
A-6.1 Retention of Records And Reports
146 ARTICLE A-7 REFERENCES
147 NONMANDATORY APPENDIX B REGULATORY ADMINISTRATIVE PROVISIONS FOR NUCLEAR REACTOR FACILITIES USING RIM PROGRAM
ARTICLE B-1 GENERAL REQUIREMENTS
B-1.1 Scope
B-1.2 Application of Code Edition
B-1.3 Application of Code Cases
B-1.4 Review by Regulatory and Enforcement Authority Having Jurisdiction at the Facility Site
B-1.5 Report Submittal
148 ARTICLE B-2 REQUIREMENTS FOR PASSIVE COMPONENTS IN THE RIM PROGRAM
149 ENDNOTES
ASME BPVC XI 2 2023
$377.00