The integrity of the reactor pressure vessel is critical to plant safety. A failure of the vessel is beyond the design basis. Therefore, the design requirements for vessels have significant margins to prevent brittle or ductile failure under all anticipated operating conditions. The early vessels in the United States were designed to meet the requirements of ASME BPVC Section VIII. The design requirements for these vessels were supplemented by special requirements based on earlier U.S. Navy design experiences. In general, the allowable stress limits were lower than vessels designed to later ASME BPVC Section III requirements. The early design codes did not include the rigorous fracture toughness requirements found in today's codes. Section VIII relied on a “fracture-safe” design approach, which sought to ensure that pressure vessels were operated in a temperature regime (upper-shelf region) where small flaws or cracks would not affect the load-carrying capacity of the structure.

Primary water stress corrosion cracking (PWSCC) of alloy 600 nickel-chromium-iron base metal and related alloys 82 and 182 weld metal has become a principal materials degradation concern for commercial pressurized water reactor (PWR) plants. Cracks and leaks have been discovered in alloys 600/82/182 materials at numerous PWR plant primary coolant system locations, including at several locations in the reactor vessels. The reactor vessel locations include top head control rod drive mechanism (CRDM) nozzles, top head thermocouple nozzles, bottom head instrument nozzles, and reactor vessel outlet nozzle butt welds. The consequences of this PWSCC have been significant worldwide with about 73 leaks through July 2013 (∼57 CRDM nozzles, 13 reactor vessel closure head thermocouple nozzles, 2 reactor pressure vessel bottom-mounted instrument nozzles, and 1 piping butt weld), many cracked nozzles and welds, expensive inspections, more than 100 heads replaced, several plants with several-month outage extensions to repair leaks, and a plant shutdown for more than 2 years due to extensive corrosion of the vessel head resulting from leakage from a PWSCC crack in a CRDM nozzle. This chapter addresses alloys 600/82/182 material locations in reactor vessels, operating experience, causes of PWSCC, inspection methods and findings, safety considerations, degradation predictions, repair methods, remedial measures, and strategic planning to address PWSCC at the lowest possible net present value cost.

The objective of this chapter is to provide some highlights on the recent issues and lessons learned by the Nuclear Regulatory Commission (NRC) for the safety related and safety significant nuclear power plant components, which utilize the ASME Boiler and Pressure Vessel Codes, Sections III and XI. These include issues related to 2011 Japan Earthquake that directly impacted the Fukushima Nuclear Plant, and also led the regulatory bodies worldwide to take a closer look at the emergency systems for any needed changes for beyond the design basis events. This chapter also provides some information on environmental assisted fatigue of nuclear components exposed to Light Water Reactor (LWR) environments, buried piping leakage and contamination issues, the issues related to High density polyethylene (HDPE) as an alternate piping material for the replacement of buried steel piping, the issues and lessons learned related to BWR steam dryers subjected to flow induced vibration loading associated with extended power uprate (EPU) conditions, the current conclusions related to quasi-laminar indications in reactor pressure vessel ring forgings, steam generator tube leaks and fluid elastic instability, and lessons learned from Operating plants experience for consideration in new nuclear power plant (NPP) construction.

Chapter 35, the functionality was initially authored by Guy Deboo for the previous two editions. This third edition has been updated by Stephen R. Gosselin who revised the discussions pertaining to and operability criteria, which address evaluations for operating plant systems, structures, or components (SSCs) found to be degraded, nonconforming, or subjected to unanalyzed conditions during nuclear plant operation. This revision discusses the methodologies and acceptance criteria applicable to these evaluations. Gosselin introduces typical SSCs that may require operability assessments and functionality evaluations and discuss methods and assessments, failure modes, functionality and operability, and as-built conditions divergent from design. He covers, with the aid of figures, tables, and references, Code requirements as well as short-term and long-term operability acceptance criteria for valves, pumps, snubbers, piping, reactor vessels, tanks, heat exchangers and supports (including component standard and linear supports as well as spring hangers), structural bolts, concrete expansion anchors, and integral welded attachments. The current practice involves a process of consensus among the regulator viewpoints; plant-specific Technical Specification (TS) requirements; and applicable Codes, Standards, rules, and other licensing-basis compliance requirements. Guy discusses the role of related agencies and committees, such as the U.S. NRC, the ASME Code Committees, and the ASME O&M Code Committees. This chapter includes basic concepts, definitions, evaluation methods, and acceptance criteria from these documents. In this chapter, the scope of SSCs is limited to mechanical systems and their components and supporting structures. Authors discuss the role of the CFR-facility TSs as they relate to the topics of this chapter. Authors provide examples of specified safety functions, operating conditions, and events to be considered for some SSCs and piping. The discussions elucidate the often complex, and sometimes non-uniform nature of operability concepts and criteria.

Chapter 5, authored by Richard W. Swayne, describes the general requirements of Section III applicable to all Construction Classes, including concrete structures and steel vessels, piping, pumps, and valves. It identifies how to classify components and describes how the jurisdictional boundaries of Section III define what is within and what is outside the scope of the Code. This chapter includes coverage of Subsection NCA, which pertains to general requirements for Divisions 1, 2, and 3 of Section III. Division 1 includes steel items such as vessels, storage tanks, piping systems, pumps, valves, supports, and core support structures for use at commercial nuclear power plants; Division 2 includes concrete reactor vessels and concrete containment vessels; and Division 3 includes requirements for the construction of containment vessels for transportation of spent nuclear fuel. The scope of Division 3 now also includes recently-published requirements for construction of storage canisters for spent nuclear fuel and spent-fuel transportation-containment vessels. Chapter 5 also explains the use of Code Editions, Addenda, and Code Cases. The requirements for design basis, design and construction specifications, and design reports are described, and the responsibilities and quality assurance program requirements of the different entities involved in nuclear power plant construction—from the Material Manufacturer to the Owner—are addressed. Requirements for ASME accreditation, application of the ASME Code Symbol Stamp, and use of Code Data Reports are described. With in-depth information, Mr. Swayne outlines the basis for exemptions, component classification, load combinations, responsibilities, Certificate of Authorization Holders and Quality System Certificate Holders. Also, Mr. Swayne provides cross-referencing to other Code Sections and Subsections, such as Sections III and XI, as well as to pertinent Regulatory Guides, such as the U.S. Code of Federal Regulations (CFR).

Chapter 11, authored by John T. Land, deals with Subsection NG ( Core-Support Structures ). This chapter provides commentary and practical examples on the materials, design, fabrication, installation, and examination requirements for core-support structures in Section III, Division 1, Subsection NG. In addition, commentary on Section XI as it applies to core-support structure repair, replacement, examination, and inspection requirements is presented. In the first edition, the 1998 Edition of the Code was used to provide examples and discussion points. In this edition, the 2001 Edition of the Code up to and including July 2003 Addenda is used to provide examples and discussion points. The objective of the Subsection NG rules is to provide a Code for the design and manufacture of structures that support the core in pressurized water reactors (PWRs) and boiling water reactors (BWRs). John indicates the subtle differences and overlaps between this Subsection and other Code Subsections. With the aid of figures, tables, and examples, John discusses important considerations in the design of core-support structures, the Owner's Design Specification, and the jurisdictional boundaries between core-support structures and reactor pressure vessels (RPVs). John explains the differences between core-support structures and internal structures, threaded structural fasteners, and temporary attachments. Discussions also include unique conditions of service; construction materials; special materials; fabrication and installation rules; examination and repair; general design rules; design by analysis; testing and overpressure protection; and examples of load combinations for core-support structures. The third edition of this chapter has been updated to the 2007 Edition of the ASME B&PV Code with new or additional commentary covering: Background on Subsection NG Development; Discussion of Typical Materials Used in CSS, IS, and TSFs; Owner's Design Specification and Design Reports; Environmental Effects; CSS Code Cases; Improvements in Subsection NG; Material Degradation Issues; Compatibility of Subsection NG with Other International Codes; Trends Towards Realistic Design Loads in Reactor Internals; and Summary of Changes to 2007 Edition of the ASME Code for CSS.

Chapter 14 describes the bases and provisions of the Code for Concrete Reactor Vessels and Containments updating to 2007 Code Edition. After a short description of the provisions for Concrete Reactor Vessels, the Chapter describes the concrete containment general environment, types of existing containments, future containment configurations, and background development including the regulatory bases of concrete containment construction code requirements. The description covers sequentially the following topics: Introduction, Concrete Reactor Vessels, Concrete Reactor Containments, Types of Containments, Future Containments, Regulatory Bases for the Code Development, Background Development of the Code, Reinforced Concrete Containment Behavior, Containment Design Analysis and Related Testing, Code Design Requirements, Fabrication and Construction, Construction Testing and Examination, Containment Structural Integrity Testing, Containment Overpressure Protection, Stamping and Reports, Containment Structure and Aircraft Impact, Containment and Severe Accident Considerations, Other Relevant Information, Summary and Conclusion. The previous editions of this Chapter were developed by John D. Stevenson, and it has been expanded by the current authors, utilizing the expertise of their respective fields. The basic format of this chapter is kept the same as in the previous editions. The updates and additional information relating to the regulatory bases for the code requirements, future containments and considerations for future revisions of the Code included in this update are based on contributions from Hansraj Ashar, Barry Scott, and Joseph Artuso.

Chapter 43, authored by Timothy J. Griesbach, covers PWR Reactor Vessel Integrity and the ASME Boiler and Pressure Vessel Code. The authors' objective is to provide an overview of the codes and regulations for prevention of brittle fracture of reactor pressure vessels. The background and bases for the original Section III, Appendix G Code requirements are discussed along with a description of the recent improvements that have been implemented in the Code in Section XI, Appendix G using more up-to-date technology. The changes and improvements are detailed such as the method for determining stress intensity factors, structural factors to account for uncertainties in the analytical methods, and material reference toughness curves. While the Code has incorporated these technical changes, the philosophy of protecting the vessel against brittle fracture has remained the same. The chapter also discusses ongoing efforts to incorporate the Master Curve approach for vessel toughness into the ASME Code, it considers areas for future improvements in the Code method for brittle fracture prevention of PWR reactor vessels, and it summarizes the aging management of PWR reactor vessel internals.

Jeffrey Gorman, Steve Hunt, Pete Riccardella authored Chapter 44 for the previous edition that has been updated by Pete Riccardella and Glenn White for this edition. They have considerable expertise and experience in handling PWR Reactor Vessel Alloy 600 and related issues confronted by the industry. Considering the extreme importance of this topic the authors have covered concerns pertinent to several ramifications of the problem. Primary water stress corrosion cracking (PWSCC) of Alloy 600 nickel-chromium-iron base metal and related Alloy 82/132/182 weld metal has become an increasing concern to commercial pressurized water nuclear power plants. Cracks and leaks have been discovered in Alloy 600/82/182 materials at a number of locations in PWR reactor vessels and other reactor coolant loop components worldwide. These locations include control rod drive mechanism (CRDM) nozzles, bottom head instrument nozzles, reactor vessel nozzle butt welds, and pressurizer nozzle welds. The consequences of PWSCC have been significant including numerous leaks, many cracked nozzles and welds, expensive inspections, more than 60 reactor vessel heads replaced, and extensive repair and mitigation activities on reactor coolant loop butt welds. A number of plants experienced months-long outage extensions to repair leaks, and one plant was down for over two years as a result of regulatory action following the detection of extensive corrosion to the vessel head resulting from a leaking CRDM nozzle. This chapter addresses Alloy 600/82/182 material locations in reactor vessels, operating experience, causes of PWSCC, inspection methods and findings, safety considerations, degradation predictions, repair methods, remedial measures, and strategic planning to address PWSCC at the lowest possible net present value cost. Recent industry and ASME Code activities to address these concerns are also discussed.

In Chapter 64 Dr. Milan Brumovsky discusses the Czech and Slovakian Codes with respect to the Nuclear Power Plants (NPPs) Jaslovske Bohunice (440 MW) in Slovakia, Dukovany (440 MW) and Temelín (1000 MW) in the Czech Republic (both in former Czechoslovakia). Dr. Brumovsky mentions the agreement between the former Czechoslovakia and Soviet Union in context of mutual cooperation in building NPPs. The author traces the Government decisions regarding an extended project of the technical standard documentation of NPPs organized by the International Economic Association “Interatomenergo” in Moscow. The association was set up to cooperate in the field of nuclear power between individual member states of the Council of Mutual Economical Co-operation (CMEA).. The entire complex of technical standard documentation ended in 1990, when GAEN finished the whole project at international level and consequently also in the Soviet Union. Dr. Brumovsky mentions that the fundamental problem of the project was a question of legal obligation to CMEA standards. Elaboration of obligatory position of state regulatory bodies among the members of the CMEA was done. This facilitated in determining the documentation of technical standards in the form of a legal-agreement. From the point of international relations, the procedure could be considered as sufficient; but from the standpoint of Czech NPPs, the effectiveness of utilizing these standards was at zero point, since effective steps were not organized to bring them into action. The CMEA rules resulted in merely upgrading of the Soviet rules and standards incorporated into new set of Soviet rules and standards issued around 1989. These rules and standards existed for service lifetime assessment of reactor components and were limited only to design and manufacturing; in very special cases these rules were for operation also but not from the lifetime evaluation point of view. Thus, assessment of defects, found during in-service inspection, has to be based on acceptance levels valid for manufacturing and on special procedures, prepared by the Nuclear Research Institute (NRI) Rez and manufacturers of components; for case by case application, these had to be accepted by the Czech State Office for Nuclear Safety (SONS). SONS requirements for Lifetime Evaluation and mentions that in 1993, the SONS initiated a project “Requirements for Lifetime Evaluation of VVER Main Components” (VVER: Water—Water Energetical Reactor is of pressurized water reactor type but designed and manufactured in accordance with former Soviet codes and rules). Within the scope of this project, preparation of regulatory requirements for lifetime evaluation of reactor components, including all aspects of integrity and degrading processes of these components, was performed. Responsibility of this project was given to the NRI Rez, which focused on reactor pressure vessel (RPV) and reactor internals and issued as a SONS document with recommendations that included Operational Safety Reports. In this document, no practical procedure for lifetime evaluation was given; only general and some detailed technical requirements for evaluation of these two components were described. Dr. Brumovsky discusses the NTD ASI Code for VVER Reactor Components. He mentions that approximately during the same time, a second activity was initiated by the Czech Association of Mechanical Engineers (ASI), which decided that a set of codes for reactor components, namely, Normative Technical Documentation (NTD) was needed for Czech nuclear industry. A plan for preparation of such codes was discussed, accepted, and put into action, details of which are presented in the chapter. Next is a discussion of the VERLIFE PROCEDURE which is a proposal for the European Union 5th Framework Programmes that was prepared and accepted with the aim to use proposals of the Section IV as the first document to be discussed, changed, upgraded, enlarged, and finally accepted. The main goal of the project was in the preparation, evaluation, and mutual agreement of a “Unified Procedure for Lifetime Assessment of Components and Piping in VVER Type Nuclear Power Plants.” The author thereon discusses the COVERS CONTINUATION. In 2005, a new project within the EU 6th Framework Programmes was opened: COVERS—VVER Safety Research that is also coordinated by the NRI. In this project, WP 4 deals with the upgrading and updating of the VERLIFE procedure to assure that the experience obtained as well as new developments will be appropriately included in the new version. Experts from nine countries are taking part in this project, in addition to VVER-operating countries such as Czech Republic, Slovak Republic, Hungary, Finland, Spain, The Netherlands, Germany, Russia, and Ukraine, as well as from EU-JRC IE (Joint Research Center—Institute of Energy in Petten, The Netherlands) and ISTC (Institute for Scientific and Technical Cooperation). Dr. Brumovsky concludes that The VERLIFE procedure is now fully accepted as a main regulatory document for lifetime assessment of VVER components in the Czech Republic and Slovakia and partially in Hungary and Finland. Negotiations are now in progress for its use in Ukraine and China. The chapter has information about several manufacturing companies in the Czech Republic, Slovakia that obtained ASME Certification for manufacturing reactor (and also nonreactor components in accordance with ASME Section VIII) components for export to other countries where ASME Codes are required. The author provides References with annotated bibliography and author's publications pertinent to this chapter. Dr. Brumovsky provides detailed information about the Structure of NTD ASI. The final version of the VERLIFE procedure in Czech was accepted as a new version of the Section IV of the NTD ASI. Czech SONS accepted NTD ASI Sections I, II, III, and IV in 2005 and recommended them for their use in the chosen safety important components in NPPs. Similarly in the Slovak Republic, Sections I and II, prepared by the Welding Institute of Slovakia in cooperation with the Welding Institute of the Czech Republic were accepted by Slovak Office for Nuclear Regulation. Structure of the Sections I, II, and III is similar to the appropriate Sections of the ASME Code Sections I, II, and III, where as the structure of Sections IV and V is fully new. The author provides detailed comparison of each of the Czech Codes with ASME B&PV Code Sections I, II, III, IV and V.

Chapter 35, authored by Guy H. DeBoo, discusses the functionality and operability criteria, which address evaluations for operating plant systems, structures, or components (SSCs) found to be degraded, nonconforming, or subjected to unanalyzed conditions during nuclear plant operation. This chapter addresses the methodology and acceptance criteria applicable to these evaluations. Guy discusses SSCs that require operability evaluation methods and assessments, failure modes, functionality and operability, and as-built conditions divergent from design. He covers, with the aid of figures, tables, and references, Code requirements as well as short- and long-term operability acceptance criteria for valves, pumps, snubbers, piping, reactor vessels, tanks, heat exchangers and supports (including component standard and linear supports as well as spring hangers), structural bolts, concrete expansion anchors, and integral welded attachments. The current practice involves a process of consensus among the regulator viewpoints; plant-specific Technical Specification (TS) requirements; and applicable Codes, Standards, rules, and other licensing-basis compliance requirements. Guy discusses the role of related agencies and committees, such as the U.S. NRC, the ASME Code Committees, and the ASME O&M Code Committees. This chapter includes basic concepts, definitions, evaluation methods, and acceptance criteria from these documents. In this chapter, the scope of SSCs is limited to mechanical systems and their components and supporting structures. Guy discusses the role of the CFR-facility TSs as they relate to the topics of this chapter. Guy provides examples of specified safety functions, operating conditions, and events to be considered for some SSCs and piping. His discussion elucidates the often complex, sometimes nonuniform nature of operability concepts and criteria.

In a PSA Level 2 for an outage state the prerequisites are quite different compared to an analysis of the power operation state or of shut-down and start-up operation. From a level 2 and source term point of view some of the main differences are: • The containment can not be assumed to be operational, except for a short time in the beginning and at the end of the outage period. • Only few safety systems or severe accident mitigating systems are in operation during most of the outage time. As an example, the containment sprays, the containment venting system and the pressure suppression system are not working for most of the outage period. • The types of fuel damage include not only overheated and melted fuel, but the fuel may also be damaged mechanically, by criticality events or by fire. • The damaged fuel may not only be situated in the reactor vessel but also in the fuel pools. In a level 2 PSA for power operation, the source term is often assessed by integral codes like MAAP, MELCOR or ASTEC. These codes are designed for accidents where the fuel is overheated and positioned in the reactor vessel. As a result, they are not easily adapted to accidents typical for outage analysis. The number of accident sequences is considerable lower than in the PSA for power operation, but it is still advantageous from both an analytical and a pedagogical point of view to use the concept of release categories. The Oskarshamn 1 Level 2 PSA for the outage period was finalized in late autumn 2005. The paper presents analysis assumptions, estimated source terms and a discussion on general results.

In the framework of a level 2 PSA for the German boiling water reactor type SWR 69, a probabilistic evaluation of low power and shutdown (LP&SD) states was performed by GRS on behalf of environmental ministry (BMU/BfS). This paper presents preliminary insights and results of the PSA for the low power and shutdown modes of this plant type. Dominant sequences are described in more detail. A detailed evaluation of the German operating experience was performed, to find events, which can lead to initiating events or, which can influence the control of accidents during shutdown and outage. Precursor events, e. g. for tube rupture at the RPV due to heavy load drops, leaks at the reactor pressure vessel (RPV) bottom and the absence of control rods during core loading with fuel elements were found in the German operating experience. Significant for this plant type is the risk contribution due to leaks inside the containment, because of the special containment design. During the outage, the containment bottom is open. Therefore the sump function is not available in case of leaks and the coolant can flood the rooms where the residual heat removal system is installed. The leakage can be returned to the reactor cavity by a one train system located in the sump of the reactor building. A finding from thermo-hydraulic analyses was, that core damage can not occur as long as the operational systems for control rod drive flushing and pump seal water are in operation. The small operational injection rate of these systems is sufficient to compensate the vaporized coolant. The calculated preliminary overall core damage probability (CDP) per outage for the considered BWR plant type in low power and shutdown modes is in the range of 5·10 −6 and about 5 times higher than the CDP calculated for full power operation. The results show the importance of event sequences during low power and shutdown operation. The safety of this plant type can be improved by plant modifications. Such modifications can be improvements of the procedures for accident control or provision of more system redundancy during outage. Highest contributors to the calculated core damage frequency yield the initiating events “loss of preferred power” and “loss of residual heat removal”. Each of these initiating events yields a core damage probability in the range of 10 −6 .

Chapter 43, authored by Timothy J. Griesbach, covers PWR Reactor Vessel Integrity and the ASME Boiler and Pressure Vessel Code. The authors's objective is to provide an overview of the codes and regulations for prevention of brittle fracture of reactor pressure vessels. The background and bases for the original Section III, Appendix G Code requirements are discussed along with a description of the recent improvements that have been implemented in the Code in Section XI, Appendix G using more up-to-date technology. The changes and improvements are detailed such as the method for determining stress intensity factors, structural factors to account for uncertainties in the analytical methods, and material reference toughness curves. While the Code has incorporated these technical changes, the philosophy of protecting the vessel against brittle fracture has remained the same. The chapter also discusses ongoing efforts to incorporate the Master Curve approach for vessel toughness into the ASME Code, it considers areas for future improvements in the Code method for brittle fracture prevention of PWR reactor vessels, and it summarizes the aging management of PWR reactor vessel internals.

Jeff Gorman, Steve Hunt and Pete Riccardella authored chapter 44. They have considerable expertise and experience in handling PWR Reactor Vessel Alloy 600 and related issues address the problems confronted by the industry. Considering the extreme importance of this topic the authors have covered concerns pertinent to several ramifications of the problem. Primary water stress corrosion cracking (PWSCC) of Alloy 600 nickel-chromium-iron base metal and related Alloy 82∕132∕182 weld metal has become an increasing concern to commercial pressurized water nuclear power plants. Cracks and leaks have been discovered in Alloy 600∕82∕182 materials at a number of locations in PWR reactor vessels worldwide. These locations include control rod drive mechanism (CRDM) nozzles, bottom head instrument nozzles and reactor vessel outlet and inlet nozzle butt welds. The consequences of this PWSCC have been significant including 72 leaks (56 CRDM nozzles, 13 thermocouple nozzles, 2 bottom mounted instrument nozzles, and one butt weld), many cracked nozzles and welds, expensive inspections, more than 60 heads replaced, several plants with several month outage extensions to repair leaks, and one plant down for over two years with extensive corrosion to the vessel head resulting from a leaking CRDM nozzle. This chapter addresses Alloy 600∕82∕182 material locations in reactor vessels, operating experience, causes of PWSCC, inspection methods and findings, safety considerations, degradation predictions, repair methods, remedial measures, and strategic planning to address PWSCC at the lowest possible net present value cost.

Chapter 11, authored by John T. Land, deals with Subsection NG (Core-Support Structures). This chapter provides commentary and practical examples on the materials, design, fabrication, installation, and examination requirements for core-support structures in Section III, Division 1, Subsection NG. In addition, commentary on Section XI as it applies to core-support structure repair, replacement, examination, and inspection requirements is presented. In the first edition, the 1998 Edition of the Code was used to provide examples and discussion points. In this edition, the 2001 Edition of the Code up to and including July 2003 Addenda is used to provide examples and discussion points. The objective of the Subsection NG rules is to provide a Code for the design and manufacture of structures that support the core in pressurized water reactors (PWRs) and boiling water reactors (BWRs). John indicates the subtle differences and overlaps between this Subsection and other Code Subsections. With the aid of figures, tables, and examples, John discusses important considerations in the design of core-support structures, the Owner's Design Specification, and the jurisdictional boundaries between core-support structures and reactor pressure vessels (RPVs). John explains the differences between core-support structures and internal structures, threaded structural fasteners, and temporary attachments. Discussions also include unique conditions of service; construction materials; special materials; fabrication and installation rules; examination and repair; general design rules; design by analysis; testing and overpressure protection; and examples of load combinations for core-support structures.

In Chapter 14, John D. Stevenson discusses the Code for concrete reactor vessels and containments. Requirements for concrete reactor vessel construction are contained in Subsection CB of Section III, Division 2. However, this Subsection of the Code is no longer being actively maintained. Requirements for concrete containment construction are contained in Subsection CC of Section III, Division 2. John distinguishes between the confinements and the containments aspects of construction, and he discusses the containments addressed by the Code that are, in fact, secondary containments and act as final barriers to radioactive releases to the environment. John deals with the interface of ASME and ACI Codes and design basis requirements for containment design and construction. There is a section that briefly describes of the construction requirements, techniques, and procedures developed by the joint ACI 359 and ASME B&PV Code Section III, Division 2, Subsection CC Committee for concrete containments. Drawing from his wealth of professional experience, John addresses the stipulations for both BWRs and PWRs, and he provides both a historical and a global perspective on this aspect of the Code.

Chapter 5, authored by Richard W. Swayne, describes the general requirements of Section III applicable to all Construction Classes, including concrete structures and steel vessels, piping, pumps, and valves. It identifies how to classify components and describes how the jurisdictional boundaries of Section III define what is within and what is outside the scope of the Code. This chapter includes coverage of Subsection NCA, which pertains to general requirements for both Divisions 1 and 2 of Section III. Division 1 includes steel items such as vessels, storage tanks, piping systems, pumps, valves, supports, and core support structures for use at commercial nuclear power plants; Division 2 includes concrete reactor vessels and concrete containment vessels; and Division 3 includes requirements for the construction of containment vessels for transportation of spent nuclear fuel. The scope of Division 3 also includes storage canisters for spent nuclear fuel and spent-fuel transportation-containment vessels, the requirements of which have not yet been published. Chapter 5 also explains the use of Code Editions, Addenda, and Code Cases. The requirements for design basis, design and construction specifications, and design reports are described, and the responsibilities and quality assurance program requirements of the different entities involved in nuclear power plant construction—from the Material Manufacturer to the Owner—are addressed. Requirements for ASME accreditation, the application of the ASME Code Symbol Stamp, and the use of Code Data Reports are described. With in-depth information, Richard outlines the basis for exemptions, component classification, load combinations, responsibilities, Certificate of Authorization Holders and Quality System Certificate Holders. Also, Richard provides cross-referencing to other Code Sections and Subsections, such as Sections III and XI, as well as to pertinent Regulatory Guides, such as the U.S. Code of Federal Regulations (CFR).