: This chapter provides a chronological overview of the development of Section XI, from its inception in 1968 to the 2015 edition. It traces the development, edition-by-edition, of important Code elements, including the philosophy behind many of the revisions. Emphasis is placed on the 2003 Addenda through 2008 Addenda because they apply to a majority of nuclear plants in the United States and in other countries. Through an extensive tabulation of Code Interpretations, the chapter also attempts to give the Code user some insights into clarification of many Section XI requirements. Section XI has Divisions 1, 2, and 3 for light-water-cooled, gas-cooled, and liquid metal-cooled reactor systems. In the more recent revisions of Section XI, feedback from operating plants has resulted in new requirements to address stress corrosion cracking and other corrosion-related mechanisms, as well as weld overlay and other piping repair techniques, and also revisions to inspection and test requirements to address a number of other current issues such as nondestructive examination procedures used by nuclear power plant designers. The 2015 Edition includes substantial revisions. Some revisions were for clarifications based on the issued interpretations, and some were for improvements in the requirements based on industry experience. The changes included the incorporation of Code Cases N-532-5 and N-805, revision to Nonmandatory Appendix D, and the application of peening for enhancing surface properties. Finally, the chapter includes a discussion on some of the issues and efforts that did not result in the final action by the Committee, such as the rules for an inservice inspection of reactor coolant pump flywheels.

: This chapter addresses ASME Code Section XI, Subsections IWE and IWL, for nuclear power plant containments. Subsection IWE provides requirements for preservice and inservice examination/inspection, repair/replacement (R/R) activities, and testing of Class MC (metal containment) pressure-retaining components and their integral attachments and R/R activities. It also provides requirements for testing of Class CC (concrete containment) pressure-retaining components and their integral attachments for boiling water reactors (BWRs) and pressurized water reactors (PWRs). Similarly, Subsection IWL provides requirements for preservice and inservice examination/inspection, R/R activities, and testing of the reinforced concrete and post-tensioning systems of Class CC components for BWRs and PWRs. Together with Subsection IWA, a comprehensive basis is provided for ensuring the continued structural and leak-tight integrity of containments in nuclear power plants. The chapter includes a brief history of the development of ASME Code requirements, the latest information on the U.S. Nuclear Regulatory Commission regulatory review and incorporation of recent ASME Code Editions and Addenda, and current listings of U.S. ‘construction-complete’ commercial nuclear power plants and their containment types and operating status. A summary of recent nuclear industry developments in advanced plant designs and a discussion of the as-yet-unrealized ‘U.S. nuclear renaissance’ are also presented, along with information on the global demand for energy and the current status of commercial nuclear power around the world. Lastly, the current Subsection IWE and IWL Commentaries, as approved and maintained by the ASME Working Group on Containment, are included.

: This chapter reflects the 2015 Edition of Section III, Division 3 of the ASME Boiler and Pressure Vessel (BPV) Code. Division 3 covers all “construction” aspects of the containment of nuclear transportation packagings, including administrative requirements, material selection, material qualification, design, fabrication, examination, inspection, testing, quality assurance, and documentation. The current scope for Division 3 includes the containments for both transportation and storage and the internal support structures for both containments. The chapter first traces the history of the ASME Subgroup NUPACK, and outlines the scope and requirements of Division 3. It then discusses the system of loading classification and the design rules based on these loadings. Division 3 loadings include Design Loadings (Design Pressure, Design Temperature, and Design Mechanical Loadings), Operating Loadings (normal, off-normal, and accident), and Test Loadings (from pressure and leak tests). The chapter also provides a brief review of the ASME stress evaluations and the associated stress acceptance criteria. Code text organization of Division 3 is then considered. The ASME BPV Code Section III, Division 3 is currently divided into three subsections: WA, General Requirements; WB, Class TC Containments; and WC, Class SC Storage Containments. Discussions also include the incorporation of strain-based acceptance criteria, and the current activities in Division 3 such as the development of new subsection WD addressing internal support structures. The chapter concludes with a summary of suggested enhancements for the future in the scope for Division 3.

: The role of fracture mechanics in ASME Section XI applications comes in the form of evaluation of indications or flaws detected during inservice inspection of nuclear components. The need to include the influence of significant plastic deformation, which may accompany crack initiation and the subsequent stable growth, has been the main driving force for the development of the field of elastic-plastic fracture mechanics (EPFM). This chapter traces the history of development of the EPFM from the 1960s and reviews the EPFM applications in Section XI Code. It discusses the three approaches considered in the application of EPFM in flaw evaluations: J -integral tearing modulus-based approach or J-T methodology, deformation plasticity failure assessment diagram (DPFAD) methodology, and R-6 methodology. The engineering estimation of the applied J integral is described, followed by a discussion on the application of the J-T approach in the evaluation of flaws in austenitic stainless steel piping and ASME Class 1 ferritic piping. The chapter describes the application of EPFM to reactor pressure vessel evaluation in three distinct ways: upper-shelf energy evaluation, Section XI flaw evaluation approach, and a probabilistic approach. It also explains the application of the Master Fracture Toughness Curve method to evaluate pressure vessel steels based on EPFM. The current activity within the ASME Section XI Code Groups related to the potential revision of ASME Code Case N-830 is finally summarized.

For the sixth edition, this chapter has been updated with revised Code requirements and Code Cases from the 2017 Code and 2019 Code Editions. The important areas of revised requirements and procedures include flaw proximity rules in IWA-3300, change in the yield strength limits from 50 ksi to 90 ksi for selected grades of SA-508 and SA-533, addition of stress intensity factor equations for circumferential OD surface flaws in Appendix A, A-3000, revised procedures for pipe flaw evaluation in Appendix C, C-4000 and C-5300, and an updated glossary terminology/definitions. Several new and revised Code Cases are also discussed to include N-513-5 (temporary acceptance of flaws in moderate energy piping systems), N-705-1 (temporary acceptance of flaws in moderate energy vessels and tanks), N-806-1 (evaluation of wall loss in buried pipe), N-848-1 (proximity rules for quasi-laminar flaws), N-869 (extension of N-513 to higher pressures for Class 2 and 3 piping systems), N-877 (characterization of multiple subsurface radially oriented planar flaws), N-889 (new reference curves for irradiated-assisted stress corrosion cracking growth for austenitic stainless steels), N-899 (weld residual stress reference curves for dissimilar metal welds), and N-890 (materials exempted from Appendix G, G-2110(b) requirements). It is important to note that additional proposal changes are being developed by the committees, some of which are discussed in 30.11. History Arthur F. Deardorff was the original author of this chapter that was updated by Mr. Deardorff in the second edition. For the third through fifth editions, Russell C. Cipolla revised the entire chapter although retaining certain portions of the previous text, equations, tables, and figures. For the current online edition, Mr. Cipolla has updated using 2017 and 2019 Editions, and additional information as appropriate.

This chapter discusses the evaluation of flaws in nuclear power plant components using the methods and criteria in ASME Code Section XI. It covers flaw acceptance based on the use of predefined acceptance standards and the use of detailed fracture mechanics evaluation. A detailed discussion of the major technical areas are provided and includes the following topics; flaw characterization; flaw acceptance standards; component flaw evaluation (reactor pressure vessel, etc.); piping flaw evaluation (for austenitic and ferritic materials); evaluation of piping wall thinning, such as wall loss by flow-assisted corrosion (FAC); and temporary acceptance of flaws in moderate energy systems. The information presented in this chapter can probably be most easily understood if it is read in conjunction with a copy of Section XI. Although Section XI strictly applies to nuclear power plant components and piping, the acceptance criteria and evaluation procedures may be applicable to the assessment of other types of pressure vessels, piping, and associated components. In addition, the criteria and evaluation procedures in Section XI are constantly being improved and expanded by Code Committee activities, both by Code Cases and by Code changes. The use of Section XI and its Code Cases for other applications of flaw inspection and evaluation may be considered, but it remains the responsibility of the user to ensure that it is appropriate.

If plant systems, structures, or components (SSCs) are found to be degraded, nonconforming, or subject to an unanalyzed condition during operation, an evaluation of their functionality or operability may be necessary to demonstrate the capability of the nuclear power generating station to continue operating safely. This evaluation establishes the ability of the SSCs to perform their assigned safety functions. This chapter discusses the methodology and acceptance criteria applicable to this evaluation. It describes typical SSCs that may require operability evaluations and also provides definitions related to operability. Examples of the operating conditions and events to consider are described, as are the evaluation methods and acceptance criteria for short- and long-term operability.

This chapter discusses the evaluation of flaws in nuclear power plant components using the methods and criteria in ASME Code Section XI [1]. It covers flaw acceptance based on the use of predefined acceptance standards and the use of detailed fracture mechanics evaluation. A detailed discussion of the major technical areas are provided and includes the following topics flaw characterization; flaw acceptance standards; component flaw evaluation (reactor pressure vessel, etc.); piping flaw evaluation (for austenitic and ferritic materials); evaluation of piping wall thinning, such as wall loss by flow-assisted corrosion (FAC); and temporary acceptance of flaws in moderate energy systems. The information presented in this chapter can probably be most easily understood if it is read in conjunction with a copy of Section XI. Although Section XI strictly applies to nuclear power plant components and piping, the acceptance criteria and evaluation procedures may be applicable to the assessment of other types of pressure vessels, piping, and associated components. In addition, the criteria and evaluation procedures in Section XI are constantly being improved and expanded by Code Committee activities, both by Code Cases and by Code changes. The use of Section XI and its Code Cases for other applications of flaw inspection and evaluation may be considered, but it remains the responsibility of the user to ensure that it is appropriate.

A Monte Carlo model of Elekta iViewGT amorphous silicon electronic portal imaging device (a-Si EPID) has been validated for pre-treatment dose verification of clinical IMRT treatment plans. The simulation employed BEAMnrc and DOSXYZnrc MC codes to predict the dose deposition in the iViewGT a-Si EPID model. The predicted dose distributions were compared to the measured dose distributions obtained from the experiment. The measured EPID images were obtained by delivering a photon beam from Elekta Precise linac to the Elekta iViewGT a-Si EPID. The a-Si EPID was used as non-transit dosimetry with no extra build-up material placed above the EPID detector. The EPID images were acquired using frame-averaging acquisition and were processed using in-house application software. The agreement between the predicted and measured dose distribution was analyzed using the gamma analysis technique with the acceptance criteria of 3%, 3 mm. The results show that the predicted dose distributions of four clinical IMRT treatment plans have a good agreement with the measurement. Three prostate IMRT plans have the average gamma pass rate more than 95% and a spinal IMRT plan has the average gamma pass rate of 94.27%. Overall, Monte Carlo simulation shows a suitability to be implemented for pre-treatment dose verification, promising a better accuracy than the existing algorithms in Treatment Planning System (TPS).

In Chapter 29, which was originally written by Arthur F. Deardorff, and updated and expanded by Russell C. Cipolla, the flaw acceptance criteria and evaluation methods specified in the 2007 Edition of ASME Section XI Code are discussed. Coverage includes the evaluation of flaws in nuclear power plant components and piping using ASME Section XI procedures. The authors discuss flaw acceptance criteria based on the use of predefined acceptance standards and of detailed fracture-mechanics evaluations of flaws. Commentary is provided on flaw characterization and acceptance standards, Class 1 vessel flaw evaluation, piping flaw evaluation (for austenitic and ferritic materials), and evaluation of piping thinned by flow-assisted corrosion. The authors discuss the background and philosophy of the Section XI approach for evaluating in-service degradation, including the rules for in-service inspection of nuclear power plant components and piping as they relate to the criteria, to determine if flaws are acceptable for continued operation without the need for repair. Drawing upon their participation in Code Committees and professional experience with both domestic and international nuclear plants, the authors discuss step-by-step procedures for the evaluation of flaws in austenitic and ferritic components and piping. The underlying philosophy of Section XI evaluation of degraded components is to provide a structural margin consistent with that which existed in the original design and construction code. Russ has expanded the chapter to describe the updated flaw evaluation procedures for piping, which were added to Section XI in 2002. Also discussed are revised flaw acceptance criteria for Class 1 ferritic vessels in IWB-3610, updated structural factors for austenitic and ferritic piping in Appendix C, and revised fatigue crack growth reference curves, along with the technical basis for these changes. Russ has also added the historical development and technical basis for Appendices E, G and K, which deal with evaluations for fracture prevention during operating plant events/conditions in the fracture-toughness transition temperature region, and at upper shelf. Further, recent Code Cases N-513 and N-705 to Section XI are described, which cover the requirements and procedures for temporary acceptance of service induced degradation in piping and vessels in moderate energy Class 2 and 3 systems. The degradation can be associated with various mechanisms (cracking, pitting, general wall thinning, etc.) and can include through-wall degradation where leakage can be adequately managed via monitoring. These Cases provide the basis for continued operation until repair can be implemented at a later time. Wherever possible, the authors cite references to published documents and papers to aid the reader in understanding the technical bases of the specified Code flaw evaluation methods and acceptance criteria. The authors also cite related Section XI requirements that are discussed in Chapters 26, 27, 28, 31, 32, 35, and 39 of this book.

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.

The aim of the shut down analysis was to produce a full Safety Analysis Report (SAR) for the shutdown and outage states including re-fueling. The SAR should contain both a deterministic and a probabilistic analysis (this is a requirement in Sweden today). The two analyses were performed in parallel with both common parts and common support activities. The probabilistic analysis should comprise both Level 1 and Level 2. The main intention of the parallel process was to form a homogeneous analysis, i.e. that equal limitations, conditions and assumptions were used. Another intention of the parallel process was to facilitate the interface between the analyses. The deterministic analysis is also becoming more realistic, which also is a reason to keep the analyses together. Project goal, content of the analysis, method descriptions and acceptance criteria were defined in common documents. It was identified that there was a need to define specific acceptance criteria for the shutdown modes. This is due to that the acceptance criteria normally are defined for full power operation with all safety systems and barriers intact and in operation (this also included the definition of plant conditions). The analysis should form a basis for the SAR, why normal PSA assumptions regarding plant operating modes and system configurations could not be accepted. The analysis would have to consider all possible situations that were stipulated by the Technical Specification (TS). The issue was that the TS should not allow a configuration that had not been analyzed. A thorough identification of different system configurations in the different phases was conducted. All initiating events were then mapped so that they could be assigned to one or more system configurations for each possible phase. The main task for the HFA (Human Factor Analysis)) was to define the initiating events caused by manual actions. The mapping of initiating events caused by manual actions was focused at: • Loss of residual heat removal • Loss of coolant • Dropped load / damage during transport • Criticality events The HFA (Human Factor Analysis) analyzed the organization for planning and conduction of the work during shutdown. The Performance Shaping Factors (PSF:s) from this analysis were used in the quantification of the human errors. The analysis was not complete when Barsebäck 2 was shutdown in 2005. This paper describes the approach of the analysis and the analysis as far as it was conducted.

Nuclear Research Institute (NRI) in Rez near Prague, Czech Republic, maintains Living Probabilistic Safety Assessment (Living PSA) program for NPP Dukovany, a four unit NPP of VVER-440 type in the Czech Republic. This project has been established as a framework for all activities in NPP Dukovany that are related to risk assessment and support for risk-informed decision making. Continuously updated PSA model is extensively used for various PSA applications including Technical Specifications (TS) assessment. However, all performed TS evaluations using PSA model were done only for a limited set of risk measures using various available criteria since the Czech regulatory body (SUJB) has not issued any specific guideline or requirement for risk-informed assessment of TS as well as has not given any associated acceptance criteria. A comprehensive guideline for risk-informed TS evaluation based on agreed methodology and prescribed criteria appeared to be necessary. The paper focuses on the process how NRI participates in development of applicable methodology for risk-informed evaluation of TS in the Czech Republic. The PSA department in NRI reviewed available approaches to risk-informed TS evaluation. Trial calculations were performed to test features of the particular methods. Based on that, the methodology proposal for TS assessment of allowed outage time (AOT) changes or adequacy has been developed in NRI. It has been passed to SUJB for comments and approval. The proposal adopted the three-tiered approach from US NRC RG 1.177. The key features of its application in NRI methodology are discussed. In addition to single-event AOT risk calculation, it recommends to assess acceptability of cumulative effect of TS changes as well as to determine adequacy of the safe state for LCO's. Great attention is paid to selection of risk configurations necessary to be evaluated. The methodology suggests using fixed criteria for conditional risk increase and average risk increase. Selection of appropriate numerical values is presented and discussed in the paper. Quantitative results of trial calculations to illustrate the presented approach are shown in the paper as well.

In Chapter 29, Arthur F. Deardorff covers the Section XI flaw acceptance criteria and evaluations using Code procedures. Coverage includes the evaluation of flaws in nuclear power plant components using the methods and criteria of Section XI. Arthur discusses flaw acceptance criteria based on use of predefined acceptance standards and of detailed fracture-mechanics evaluations of flaws. Commentary is provided on flaw characterization and acceptance standards, reactor vessel flaw evaluation, piping flaw evaluation (for austenitic and ferritic materials), and evaluation of piping thinned by flow-assisted corrosion. Arthur discusses the background and philosophy of the Section XI approach for evaluating inservice degradation, including the rules for inservice inspection of nuclear power plant components as they relate to the criteria, to determine if flaws are acceptable for continued operation while they remain in the components. Drawing upon his participation in Code Committees and professional experience with both domestic and international nuclear plants, Arthur discusses step-by-step procedures for the evaluation of flaws in austenitic and ferritic components. The underlying philosophy of Section XI evaluation of degraded components is to provide a safety margin consistent with that which existed in the original component design and construction. The initial development was for Class 1 reactor-coolant pressure-boundary components, from which the components of other Classes have evolved. Arthur cites related Section XI requirements that are discussed in Chapters 26, 27, and 28 of this book.

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.

Acceptance or behavior criteria for mechanical components are typically given in terms of allowable stresses as a function of both specified minimum yield and ultimate strength (stress). These allowable stresses have been developed as a function of the experience with pressure vessel and piping failures over the past 100 years of service. It is expected that over the next 100 years, allowable stress criteria will gradually give way to more direct measures of failure probability, as discussed in Chapter 2.

Most U.S. power reactors were built with one concept in mind: to produce electricity as efficiently, economically, and safely as possible. Plants were not designed to facilitate simple and efficient deconstruction. Since decommissioning was not planned by the original construction team, dismantling is a completely new process that must be learned for each unique facility. Fortunately, the process is becoming more predictable and manageable because of experience gained to date. Dismantling is probably the phase of decommissioning with the strongest visual impact. A major portion of the project involves dismantling, removal, and size reduction of concrete structures and metal components, such as vessels, pipes, conduit, and structural steel. Generally, the metal and concrete components have radioactive, organic, chemical, or heavy metal contamination, thereby presenting potential exposure hazards. The site and facility characterization data will show which and how much of those materials must be dismantled and removed. Dismantling involves decontamination, removal of components and structures, packaging of wastes, transport of packages, and disposal in a controlled burial facility. But the planning should be performed in reverse order. Once characterization has identified the waste streams, the next step is to contact the waste disposal facility to learn and understand its waste acceptance criteria, including permissible package/container radioactive inventory (curies and dose), package size and weight requirements, and documentation. If the facility will not accept the waste form or content, alternative measures must be taken. When the package sizes, weights, and radioactivity limits are selected, the mode of transport and number of shipments may be determined — truck, rail, or barge. The planner may then select the most cost-effective technology to segment piping, equipment, and structures to fit into the selected packages. The segmentation/removal production rates (tons/day) for each potential technique evaluated will determine the number of shipments per day or week and thereby set the overall schedule. Adjustments might be made as experience demonstrates improved productivity with other technologies. This chapter will help the decommissioning planner choose appropriate technology for specific applications. Current dismantling, removal, and size reduction technologies include mechanical saws, circular cutters, abrasive cutters, diamond wire, explosive cutting, plasma arc torch, oxyacetylene torch, arc saw, abrasive water jet, and hydraulic shears. Backhoe hydraulic rams, conventional wrecking techniques, and explosives are used to dismantle large, thick concrete structures. The use of explosives requires a certified blasting expert for safety.