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
1	FIGURES TABLES
5	TABLE OF CONTENTS
10	LIST OF SECTIONS
11	INTERPRETATIONS CODE CASES
12	FOREWORD
14	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
15	SUBMITTAL OF TECHNICAL INQUIRIES TO THE BOILER AND PRESSURE VESSEL STANDARDS COMMITTEES
18	PERSONNEL
39	PREFACE TO SECTION XI INTRODUCTION GENERAL INSERVICE TESTING OF PUMP AND VALVES OWNER RESPONSIBILITIES
40	DUTIES OF THE AUTHORIZED NUCLEAR INSERVICE INSPECTOR
41	ORGANIZATION OF SECTION XI 1 DIVISIONS 2 ORGANIZATION OF DIVISION 1 2.1 SUBSECTIONS 2.2 ARTICLES
42	2.3 SUBARTICLES 2.4 SUBSUBARTICLES 2.5 PARAGRAPHS 2.6 SUBPARAGRAPHS 3 ORGANIZATION OF DIVISION 2
43	4 REFERENCES
45	SUMMARY OF CHANGES
46	LIST OF CHANGES IN RECORD NUMBER ORDER
47	CROSS-REFERENCING AND STYLISTIC CHANGES IN THE BOILER AND PRESSURE VESSEL CODE
49	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
50	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
51	RIM-1.9 REFERENCED STANDARDS AND SPECIFICATIONS RIM-1.9-1 Referenced Standards and Specifications
52	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)
53	RIM-2.4 PLANT AND SSC RELIABILITY TARGET ALLOCATION RIM-2.4.1 Plant-Level Risk and Reliability Targets RIM-2.4.2 SSC-Level Reliability Target RIM-2.4.3 Scope, Level of Detail, and Technical Adequacy of the PRA RIM-2.5 IDENTIFICATION AND EVALUATION OF RIM STRATEGIES
54	RIM-2.5.1 Identification of RIM Strategies 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 Inspection Interval
55	RIM-2.7.3 Preservice Inspection RIM-2.7.4 Design Requirements for RIM RIM-2.7.5 Leak Detection System Requirements for RIM
56	RIM-2.7.6 Examination and Inspection Requirements for RIM
57	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
58	RIM-2.9.2 Surface Examination RIM-2.9.3 Volumetric Examination
59	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
60	ARTICLE RIM-3 ACCEPTANCE STANDARDS RIM-3.1 EVALUATION OF EXAMINATION RESULTS AND ACCEPTANCE STANDARDS
61	ARTICLE RIM-4 REPAIR/REPLACEMENT ACTIVITIES RIM-4.1 SCOPE RIM-4.2 LEAK TEST REQUIREMENTS AFTER A REPAIR/REPLACEMENT ACTIVITY RIM-4.2.1 Test Boundaries RIM-4.2.2 Gas Leak Test RIM-4.2.3 Liquid Leak Test
62	RIM-4.2.4 Volumetric and Surface Examination RIM-4.2.5 Exemptions RIM-4.3 RESPONSIBILITIES RIM-4.4 CORRECTIVE ACTION RIM-4.5 RECORDS
63	ARTICLE RIM-5 SYSTEM LEAK MONITORING AND PERIODIC TESTS RIM-5.1 SCOPE RIM-5.2 LEAKAGE MONITORING RIM-5.2.1 General RIM-5.2.2 Periodic Leak Test RIM-5.3 CORRECTIVE ACTION RIM-5.4 RECORDS
64	ARTICLE RIM-6 RECORDS AND REPORTS RIM-6.1 SCOPE RIM-6.2 REQUIREMENTS RIM-6.2.1 Owner’s Responsibilities RIM-6.2.2 Owner 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 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
65	RIM-6.3.5 Repair/Replacement Activity Records
66	ARTICLE RIM-7 GLOSSARY RIM-7.1 TERMS AND DEFINITIONS
68	MANDATORY APPENDIX I RIM DECISION FLOWCHARTS FOR USE WITH THE RIM PROGRAM ARTICLE I-1 FLOWCHARTS I-1.1 GENERAL
69	I-1.1-1 Inputs to the RIMEP for NPP Owner’s RIM Program Development
70	I-1.1-2 RIM Program Development and Integration
71	I-1.1-3 Process for Identifying the SSCs to Be in MANDE Program
72	I-1.1-4 Selection of Strategies for SSCs to Meet Reliability Targets
73	I-1.1-5 Upper Half Shows Input to MANDEEP for Developing MANDE Specification and Lower Half Shows Process for Evaluating if Division 1 Requirements Meet MANDE Specifications
74	I-1.1-6 Select, Develop, and Validate Performance Demonstration Approach to Meet SSC Reliability Target
76	MANDATORY APPENDIX II DERIVATION OF COMPONENT RELIABILITY TARGETS FROM PLANT SAFETY REQUIREMENTS ARTICLE II-1 GENERAL REQUIREMENTS II-1.1 SCOPE II-1.2 ADEQUACY OF THE PRA II-1.3 PROCEDURE OVERVIEW
77	ARTICLE II-2 RELIABILITY TARGET DERIVATION II-2.1 PLANT-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 PLANT-LEVEL RISK II-2.6 DETERMINATION OF RELIABILITY TARGETS
78	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
79	III-1.1-1 Guide for Completing Form OAR-1
80	MANDATORY APPENDIX IV MONITORING AND NDE QUALIFICATION ARTICLE IV-1 INTRODUCTION IV-1.1 SCOPE IV-1.2 METHODS IV-1.3 OWNER’S REQUIREMENTS
82	ARTICLE IV-2 PROCEDURES, EQUIPMENT, AND PERSONNEL REQUIREMENTS IV-2.1 BASIC QUALIFICATION (FIGURES I-1.1-1 THROUGH I-1.1-6) IV-2.2 METHOD/TECHNIQUE PERSONNEL-SPECIFIC QUALIFICATIONS
83	ARTICLE IV-3 RELIABILITY-BASED QUALIFICATION OF MONITORING AND NDE (MANDE) METHODS AND TECHNIQUES IV-3.1 GENERAL IV-3.2 DETERMINATION OF THE QUALIFICATION REQUIREMENTS IV-3.3 QUALIFICATION PROCESS
85	ARTICLE IV-4 PERFORMANCE DEMONSTRATIONS FOR MANDE PERSONNEL (FIGURE I-1.1-6) IV-4.1 GENERAL IV-4.2 PERFORMANCE DEMONSTRATION FOR PERSONNEL FOR MONITORING METHODS IV-4.3 PERFORMANCE DEMONSTRATION FOR NDE PERSONNEL
86	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
87	MANDATORY APPENDIX V CATALOG OF NDE REQUIREMENTS AND AREAS OF INTEREST ARTICLE V-1 TABLES V-1.1 GENERAL V-1.1-1 Examination Category A, Pressure-Retaining Welds in Reactor Vessels
88	V-1.1-2 Examination Category B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels V-1.1-3 Examination Category D, Full-Penetration Welded Nozzles in Vessels
89	V-1.1-4 Examination Category F, Pressure-Retaining Dissimilar Welds in Vessel Nozzles
90	V-1.1-5 Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter
91	V-1.1-6 Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) or Less in Diameter
92	V-1.1-7 Examination Category J, Pressure-Retaining Welds in Piping
93	V-1.1-8 Examination Category K, Welded Attachments for Vessels, Piping, Rotating Equipment, and Valves V-1.1-9 Examination Category L-2, Pump Casings; Examination Category M-2, Valve Bodies
94	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 V-1.1-11 Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings V-1.1-12 Examination Category P, All Pressure-Retaining Components
95	V-1.1-13 Examination Category F-A, Supports
96	MANDATORY APPENDIX VI RELIABILITY AND INTEGRITY MANAGEMENT EXPERT PANEL (RIMEP) ARTICLE VI-1 OVERVIEW VI-1.1 RESPONSIBILITIES AND QUALIFICATIONS OF RIMEP
97	MANDATORY APPENDIX VII SUPPLEMENTS FOR TYPES OF NUCLEAR PLANTS ARTICLE VII-1 SUPPLEMENT FOR LIGHT WATER REACTOR–TYPE PLANTS VII-1.1 SCOPE VII-1.2 RIM PROGRAM — DAMAGE DEGRADATION ASSESSMENT VII-1.3 ACCEPTANCE STANDARDS
98	VII-1.2-1 Degradation Mechanism Attributes and Attribute Criteria (LWR)
105	VII-1.3.3-1 Acceptance Standards
106	VII-1.4 ACCEPTANCE STANDARDS FOR SPECIFIC EXAMINATION CATEGORIES
111	VII-1.5 ANALYTICAL EVALUATION OF PLANAR FLAWS
114	VII-1.6 ANALYTICAL EVALUATION OF PLANT OPERATING EVENTS
115	ARTICLE VII-2 SUPPLEMENT FOR LIQUID METAL REACTOR–TYPE PLANTS
116	ARTICLE VII-3 SUPPLEMENT FOR HIGH-TEMPERATURE GAS REACTOR–TYPE PLANTS 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
117	VII-3.2-1 Degradation Mechanism Attributes and Attribute Criteria
122	VII-3.3.2 Supplemental Examinations VII-3.3.3 Acceptance Standards
123	VII-3.3.4 Characterization VII-3.3.5 Acceptability VII-3.4 ACCEPTANCE STANDARDS FOR SPECIFIC EXAMINATION CATEGORIES VII-3.4.1 Standards for Examination Categories A and B of Pressure-Retaining Welds in Reactor Vessel and Other Vessels VII-3.3.3-1 Acceptance Standards
124	VII-3.4.2 Standards for Examination Category D for Full Penetration Welds of Nozzles in Vessels
125	VII-3.4.3 Standards for Examination Category F for Pressure-Retaining Dissimilar Metal Welds in Vessel Nozzles and Category J for Pressure-Retaining Welds in Piping
126	VII-3.4.4 Standards for Examination Category G-1 for Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter VII-3.4.5 Standards for Examination Category K for Welded Attachments for Vessels, Piping, Pumps, and Valves
127	VII-3.4.6 Standards for Examination Category G-1 for Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter and Examination Category G-2 for Pressure-Retaining Bolting 2 in. (50 mm) and Less in Diameter VII-3.4.7 Standards for Examination Categories L-2 and M-2 of Rotating Equipment Casings and Valve Bodies VII-3.4.8 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
128	VII-3.4.9 Standards for Examination Category P, All Pressure-Retaining Components VII-3.4.10 Standards for Examination Category O for Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings VII-3.4.11 Standards for Examination Category F-A for Supports
129	VII-3.5 ANALYTICAL EVALUATION OF PLANAR FLAWS VII-3.5.1 Acceptance Criteria for Ferritic Steel Components 4 in. (100 mm) and Greater in Thickness
130	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
131	VII-3.5.4 Evaluation Procedure and Acceptance Criteria for Reactor Vessel Head Penetration Nozzles
132	VII-3.6 ANALYTICAL EVALUATION OF PLANT 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 Plant Fatigue Assessments
133	ARTICLE VII-4 SUPPLEMENT FOR MOLTEN SALT REACTOR–TYPE PLANTS
134	ARTICLE VII-5 SUPPLEMENT FOR GENERATION 2 LWR–TYPE PLANTS
135	ARTICLE VII-6 SUPPLEMENT FOR FUSION MACHINE–TYPE PLANTS
136	NONMANDATORY APPENDIX A ALTERNATE REQUIREMENTS FOR NDE AND MONITORING ARTICLE A-1 GENERAL A-1.1 SCOPE A-1.2 METHODS A-1.3 RESPONSIBILITIES
137	A-1.2-1 Logic Flow Diagram of the Process
138	ARTICLE A-2 PROCEDURE 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
139	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-3.5 DETERMINISTIC APPROACH — MARGIN ASSESSMENT
140	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
141	A-4.6 PROBABILISTIC APPROACH A-4.7 DETERMINISTIC APPROACH
142	ARTICLE A-5 PROCEDURE FOR STRUCTURAL RELIABILITY EVALUATION FOR PASSIVE COMPONENTS A-5.1 GENERAL REQUIREMENTS A-5.2 RELIABILITY EVALUATION A-5.2.1-1 Reliability Evaluation Procedure
143	A-5.3 FAILURE SCENARIO SETTING A-5.3.1-1 Failure Scenario Setting Procedure
144	A-5.4 MODELING A-5.4.1-1 Modeling Procedure
145	A-5.5 RELIABILITY CALCULATION
146	ARTICLE A-6 RECORDS AND REPORT A-6.1 RETENTION OF RECORDS AND REPORTS
147	ARTICLE A-7 REFERENCES
148	NONMANDATORY APPENDIX B REGULATORY ADMINISTRATIVE PROVISIONS FOR NUCLEAR PLANTS 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 AUTHORITIES HAVING JURISDICTION AT THE PLANT SITE B-1.5 SUMMARY OF REPORT SUBMITTAL
149	ARTICLE B-2 REQUIREMENTS FOR PASSIVE COMPONENTS IN THE RIM PROGRAM B-2.1 REVIEW BY REGULATORY AND ENFORCEMENT AUTHORITIES HAVING JURISDICTION AT THE PLANT SITE
150	ENDNOTES

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ASME	2021	151


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Standard Title	ASME BPVC – XI – 2 -2021 BPVC Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants
Published Code	ASME
Publication Date	2021
Pages Count	151

ASME BPVC XI 2 2021

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