Nuclear Engineering and Technology for the 21st Century - Monograph Series Jovica Riznic, Series Editor In this monograph, the authors provide an overview of Canada’s NF capability in addition to general aspects of the nuclear forensics that is useful for both nuclear forensic practitioners and for countries that are signatories to the Nuclear Non-proliferation treaty in establishing their NF capability. The authors summarize challenges that first responders face at crime scenes involving RN materials. Next, they describe the RN materials from the uranium fuel cycle in Canada that are most relevant to NF and a summary of their use and construction, possible NF signatures, and how those signatures were used in the Canadian exercise. Then there is an overview of selected NF laboratory methods, along with an important example of radiochronometric measurements. The importance of Certified Reference Materials specific for NF is described, along with the Canadian perspective. The Laboratory Network chapter provides the Canadian experience in establishing the nuclear forensic capability, while Nuclear Forensic Library chapter describes in general the need and establishment of the NNFL, with some details on the Canadian project. In the final chapter, the authors discuss the Canadian experience in using its nuclear forensic capability in two international exercises.e Nuclear Engineering and Technology for the21st Century monograph series provides currentand future engineers, researchers, technicians and other professionals and practitioners with practical, concise but key information concerning the nuclear technologies from areas of medical applications, mining, processing and manufacturing, environmental monitoring to safe and energy-efficient plant operationand electricity generation. Each monograph will provide a well-rounded and definitive state-of-the-art review of its subject, with a focus on applied research and development, and best industry practices, processes and related technological applications. The series is envisaged as a collection of concise monograph publications, which can stand as the most authoritative source of information on current state of a topic, application or discipline. This frontmatter contains the series editors’ preface, contents, preface, abstract and is available for free viewing by clicking on the PDF icon. eBook keywords: nuclear technologies, nuclear forensics, nuclear security, non-proliferation, nuclear fuel cycle, nuclear fuel manufacturing, radioactive and nuclear material, sealed radioactive sources, radiochronometry, uranium ore concentrates. Nuclear forensics (NF) is by no means new. It is as old as the first use of nuclear materials in weapons. For example, in 1944 the U.S. Air Force searched for evidence of radioactive xenon gas ( 133 Xe) in the atmosphere to monitor the German nuclear program [1]. Such monitoring activities are being practiced both internationally through the framework of the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) [2] and nationally by many countries across the world. For example, in September 2017, South Korea’s radionuclide monitoring system recorded and reported the detection of 133 Xe subsequent to nuclear bomb tests allegedly performed by North Korea (see [3]). Following a catastrophic event, such as a “dirty bomb” explosion, the crime scene may have victims, weapons, parts of the “dirty bomb,” as well as RN materials in various degrees of dispersion (intact or distributed). The primary task of the first responders and law enforcement officials would be to assess and secure the crime scene and control the evacuation of the wounded. These are extremely complex tasks for which the first responders are extensively trained. Of paramount importance is the safety of the first responders with respect to radiation hazards while tackling the management of a radiological crime scene. Therefore, the responding law-enforcement and paramedic staff require adequate training in the full assessment of a radiological crime scene for RN materials to ensure the risks due to radiation hazards are minimized and the safety of the persons operating within the environment maximized. Uranium is a naturally occurring element and its traces can be found virtually everywhere. Canada is responsible for 20% of the global supply [40] of uranium thus making it a valuable commodity of the Canadian economy. Natural uranium is comprised mainly of 238 U (99.274%) with small fractions of fissile 235 U (0.72%), and 234 U (0.005%). In addition to mining, Canada plays a dominant role in the processing of natural uranium ore from domestic and foreign sources. Canadian nuclear power plants, CANDU (Canada Deuterium Uranium) reactors, use natural uranium as a fuel and heavy water (deuterium oxide) as a moderator. Sealed sources are radioactive sources enclosed in a sealed capsule or bonded onto a protective cover. The sealed capsules must comply with the international standards to contain and prevent the leak of the radionuclide under extreme mechanical, temperature and pressure conditions [97]. Most often, sealed radioactive sources are contained in radiation devices, which are used in a variety of applications such as medical, industrial, academic and agriculture [98]. The speed and accuracy of the analysis of the nuclear forensic samples are of the essence in providing timely information to the investigative authorities, thus aiding their efforts to learn all aspects of the interdicted material in order to prevent, if possible, its use in terrorist activities. The NRC is the National Metrological Institute (NMI) for Canada and is one of the NMIs in an international family whose purpose is to establish, maintain, improve and disseminate standards. One component of the dissemination of standards is the production of Certified Reference Material (CRM) whose fundamental role is primarily the establishment of traceability of a measured value to a primary unit of measurement as defined by the International System of Units (SI). The NMIs engage each other in order to validate CRMs through inter-laboratory comparisons thereby ensuring confidence in the reliability of measurement results. Equally important to the certified values in a given CRM is the uncertainty attributed to those values, which are of the highest metrological value and are ideally and typically lower than the measurement capability of the end users. RN materials have chemical, isotopic, and physical characteristics which provide important clues in terms of their age and origin. The ability to characterize and categorize RN material allows NF to assist investigations related to a nuclear security incident. Furthermore, traditional forensic analysis of other evidence materials, which may be contaminated with RN material, also plays an important role in a nuclear forensic investigation. Thus, NF involves several areas of expertise, including elemental and isotopic analyses using mass spectrometry or radiometric measurements, non-destructive imaging techniques such as X-ray radiography, physical characterization of RN materials in various forms, and traditional forensics. NF involves gathering relevant information by analysing radioactive and nuclear (RN) materials, or evidence contaminated with RN materials, and interpreting that information to answer specific questions posed by law enforcement investigation about a nuclear security event. Given that the majority of nuclear forensic signatures are comparative, the interpretation often requires comparison with known data on the RN materials. As part of the initiative to establish a sustainable capability, the national Nuclear Forensic Laboratory Network needs to participate in both national and international exercises, including operational exercises geared towards laboratory network implementation. The operational exercise may be preceded by a series of lab inter-comparison studies focused on validating the technical aspects of the network’s analytical capabilities. These exercises should involve the end-user community, and should cover the entire “life cycle” of an event, from the incident itself to evidence collection, analysis, and data interpretation to assist the investigation. Other exercises may take the form of table-top or limited-scale, focused exercises. This back matter contains the Glossary and Abbreviations & Bibliography

An illustrative guide to the analysis needed to achieve a safe design, Stress in ASME Pressure Vessels, Boilers, and Nuclear Components offers a revised and updated edition of the classic text, Design of Plate and Shell Structures. This important resource offers engineers and students a text that covers the complexities involved in stress loads and design of plates and shell components in compliance with pressure vessel, boiler, and nuclear standards. The author covers the basic theories and includes a wealth of illustrative examples for the design of components that address the internal and external loads as well as other loads such as wind and dead loads. The text keeps the various derivations relatively simple and the resulting equations are revised to a level so that they can be applied directly to real-world design problems. The many examples clearly show the level of analysis needed to achieve a safe design based on a given required degree of accuracy. Written to be both authoritative and accessible, this important updated book: Offers an increased focus on mechanical engineering and contains more specific and practical code-related guidelines Includes problems and solutions for course and professional training use Examines the basic aspects of relevant theories and gives examples for the design of components Contains various derivations that are kept relatively simple so that they can be applied directly to design problems Written for professional mechanical engineers and students, this text offers a resource to the theories and applications that are needed to achieve an understanding of stress loads and design of plates and shell components in compliance with pressure vessel, boiler, and nuclear standards. This front matter contains the contents, series preface, acknowledgment and is available for free viewing by clicking on the PDF below. eBook keywords: pressure vessel, stress loads, shell components, boiler standards, nuclear standards, wind loads, dead loads, design of components, design of plates, stress limits. All thin cylindrical shells, spherical and ellipsoidal heads, and conical transition sections are generally analyzed and designed in accordance with the general membrane theory of shells of revolution. These components include those designed in accordance with the ASME pressure vessel code (Section VIII), boiler code (Section I), and nuclear code (Section III). Some adjustments are sometimes made to the calculated thicknesses when the ratio of radius to thickness is small or when other factors such as creep or plastic analysis enter into consideration. The effect of these factors is discussed in later chapters, whereas assumptions and derivation of the basic membrane equations needed to analyze shells of revolution due to various loading conditions are described here. The quantity N φ r 2 sin 2 φ in Eq. (1.12), when multiplied by 2 π , represents the total applied force acting on a structure at a given parallel circle of angle φ . Hence, for complicated geometries, the value of N φ in Eq. (1.12) at any given location can be obtained by taking a free-body diagram of the structure. The value of N θ at the same location can then be determined from Eq. (1.11). This method is widely used in designing pressure vessels, flat bottom tanks, elevated water towers (Figure 2.1), and other similar structures. Example 2.1 illustrates the application of this method to the design of a water-holding tank. The ASME nuclear code paragraphs NC-3932 and ND-3932 have various equations and procedures for designing components by the free-body method. This method is also useful in obtaining an approximate design at the junction of two shells of different geometries. A more accurate analysis utilizing bending moments may then be performed to establish the discontinuity stresses of the selected members at a junction if a more exact analysis is needed. The membrane stress calculated from Eq. (1.27) tends to underestimate the actual stress in a heavy wall cylinder, Figure 3.1, with an inside radius to thickness ratio of less than 10. This is due to the nonlinear stress distribution that develops across the wall of a thick cylinder when an internal pressure is applied. Cylindrical shells in boilers, pressure vessels, and nuclear components are generally designed for internal pressure as well as axial loads and external pressure. Tensile stresses due to internal pressure and axial forces are discussed in Chapters 1 through 3. Compressive stresses due to axial compressive loads and external pressure are discussed here. The required thickness of a thin spherical shell based on membrane theory is obtained from Eq. (1.13). The buckling of spherical shells under external pressure has been investigated by numerous researchers. Von Karman and Tsien (1939) developed a solution that fits experimental data very closely. Many structures such as boiler casings (Figure 7.1), submarine bulkheads, ship and barge hulls, ducts (Figure 7.2), aircraft components, and machine parts are designed in accordance with the general bending theory of plates. In this chapter, the pertinent equations of the bending theory of plates are developed, and examples are solved to demonstrate its applicability. These equations are applicable to thin plates subjected to small deflections. The majority of industrial applications encountered by the engineer fall under this category. Other theories dealing with thin plates with large deflections such as diaphragms, thick plates such as some tubesheets in heat exchangers, and composite and laminated plates in some aircraft components are beyond the scope of this book. Many of the references given at the end of the book delve into such theories, and the interested reader is encouraged to read them. Circular plates are common in many structures such as nozzle covers, end closures in pressure vessels, and bulkheads in submarines and airplanes. The derivation of the classical equations for lateral bending of circular plates dates back to 1828 and is accredited to Poisson. He used polar coordinates to transfer the differential equations for the bending of a rectangular plate to circular plates. The first rigorous solution of the differential equation of circular plates for various loading and boundary conditions was made around 1900 and is credited to Love (1944). The analysis of rectangular plates having concentrated loads, free edges, or column supports is exceedingly complicated when using the classical theory discussed in Chapters 7 and 8. Similar difficulty is encountered in circular plates with concentrated loads as well as triangular plates with various loading and boundary conditions. The analysis commonly used to solve such cases is based on the plastic yield line theory, which is discussed in this chapter. When a thin elastic plate is subjected to compressive in-plane axial loads, in conjunction with small applied lateral loads or imperfections in the plate, the in-plane deflections increase gradually with an increase in the applied loads up to a certain critical point. Beyond this point a slight increase in axial loads causes a large and sudden increase in the deflection. This phenomenon, called buckling, is the subject of this chapter for circular and rectangular plates. A more comprehensive treatment of this subject is given by Timoshenko and Gere (1961), Bloom and Coffin (2001), Szilard (1974), and Iyengar (1988). The finite element method is a powerful numerical tool for calculating stress in complicated shell and plate structures that are difficult to analyze by classical plate and shell theories. The method consists of subdividing a given domain into small elements connected at the nodal points as shown in Figure 11.1. The mathematical formulation consists of combining the governing equations of each of the elements to form a solution for the domain that satisfies the boundary conditions. The approximations associated with finite element solutions depend on many variables such as the type of element selected, the number of elements used to model the domain, and the boundary conditions. This back matter contains the appendices, references, answers to selected problems, and index

The Nuclear Engineering and Technology for the 21st Century monograph series provides current and future engineers, researchers, technicians and other professionals and practitioners with practical, concise but key information concerning the nuclear technologies from areas of medical applications, mining, processing and manufacturing, environmental monitoring to safe and energy-efficient plant operation and electricity generation. In using risk-informed approaches for ensuring safety of operating nuclear power plants (NPPs), risk importance measures obtained from probabilistic risk assessments (PRAs) of the plants are integral elements of consideration in many cases. Obtaining these measures in appropriate forms is helpful for decision makers and can facilitate the use of risk information. This front matter contains the acknowledgement, series editors’ preface, table of contents, and abstract. eBook keywords: nuclear, nuclear power plant, NPPs, risk, probabilistic risk assessment, PRAs, common-cause failure, CCF, safety analysis, energy systems, SSC. In using risk-informed approaches for ensuring safety of operating nuclear power plants (NPPs), risk importance measures obtained from probabilistic risk assessments (PRAs) of the plants are integral elements of consideration in many cases. Obtaining these measures in appropriate forms is helpful for decision makers and can facilitate the use of risk information. In order to be able to “measure” the importance of a particular feature or component which has certain safety function in any technical system (NPP included), an appropriate technical definition needs first to be established for a quantitative measure of the system’s “safety”. One of the, arguably, best ways is to present it through its inversion, the “risk”, which is, in engineer’s terms, quantitatively defined by the famous “risk curve” representing the probability (or frequency) of exceedance as a function of magnitude of consequences, Figure 2-1. This presentation corresponds, for example, to the complementary cumulative distribution function used to depict the risk in NUREG-1150. Traditional view on “importance” of a particular SSC was associated with its role (function) in ensuring appropriate safety margins (and, hence, low undesirable consequences) in the case of anticipated or credible occurrences. The “importance” of an SSC was related to its quality requirements resulting in a set of deterministic rules and principles for the classification of SSCs into quality groups such as “A”, “B”, “C” and “D” (“Class 1”, “Class 2”, “Class 3” and “Non-Class”). For example, the highest importance with regard to ensuring all safety margins concerning the cooling of the reactor core was given to the integrity of the primary system. Hence, the components of the reactor coolant pressure boundary were assigned Quality Group “A” (or “Class 1”). In the case of primary system integrity failure, ensuring the safety margins relies on the systems with functions such as emergency core cooling and containment heat removal. Thus, those systems were assigned Group “B” (or “Class 2”); supporting systems, such as cooling water systems, to the systems of Group “B” were assigned Group “C” (or “Class 3”), etc. It can be said that the idea for risk importance measures originates from the concept of traditional sensitivity analysis, which might not necessarily be related to the risk (reliability) studies. One type of sensitivity analysis is to change one of the parameters and see how much effect it makes on the result. A systematic approach would be to calculate the fractional change in result for a fractional change in a variable or parameter. At the beginning, we introduce two basic terms which will be used in all considerations to come. Both of them are “events”: Failure of component by a considered failure mode, when challenged. (This failure mode is presented in a PRA model by specifically defined single basic event.) Occurrence of specified top event representing certain damage state (e.g. reactor core damage). Many times, failure of a component is in fault trees represented by some kind of a logic structure. In other words, it is provided as some kind of a logic function defined over a set of basic events representing particular failure modes. In this section we continue consideration of component-level importance measures for a component which is not involved in any CCF group. In Section 6 the component of concern was, wherever it appears in the PRA fault tree structure, always presented by the same FT module (sub-tree). Up to now we have considered importance measures for a component represented by a FT module (sub-tree) involving a group of basic events (Section 6) or, even, a component represented by multiple different FT modules which appear in mutually exclusive sequences of failures (Section 7). However, none of the underlying basic events was involved in any common cause failure (CCF) group of basic events. (In other words, none of those failures or failure modes was shared by another components.) We will now broaden the discussion to the component-level risk importance measures for the components which are involved in a CCF group (which is, many times, the case with most of components in a PRA model). We start, in this section, with discussing the impact of CCF on risk importance of component which is presented in the FT structure by a single basic event. In many cases a particular component in a PRA model may be represented by several basic events (each related to specific failure mode) with each of them being involved in some CCF group of basic events. A demonstrative example application was carried out using a simplified PRA model to calculate the component-level risk importance measures from the PRA-calculated risk importance measures of the basic events relating to the components. The simplified PRA model used for the demonstrative examples is presented in Appendix B.1. It is based on a pressurized water reactor and is built on RiskSpectrum® PRA computer code. It is pointed out that this model was developed for demonstration of risk importance measures only and should not be related to actual NPP risk. As discussed before, risk importance measures are used in many applications to address issues relating to regulation and operation of NPPs. As being discussed in this monograph, these measures provide useful inputs and make the PRA models of NPPs present results that are directly applicable for the issues to be addressed. In these applications, the information obtained from these measures should be viewed in the context of their limitations some of which are discussed here. The last two points (i.e. 6 and 7) provide a discussion on some aspects and limitations of methods specifically considered in previous sections in this monograph. In this monograph, we have presented the risk importance measures calculated in the PRAs of NPPs and used in many applications. We have discussed the derivations of these measures and have presented relations among them using basic probabilistic principles. We have shown how each measure correlates with others and knowledge of one measure can be used to obtain other measures. This back matter contains the appendices, about the authors, and index

This monograph summarizes the major developments on nuclear reactor thermal-hydraulics over the last fifty years, primarily for the water-cooled reactors, and provides a direction for the future thermal-hydraulic developments for water-cooled, including small modular reactors or SMR, and Generation IV reactors. This includes discussion on the steady-state reactor thermal hydraulics including subchannel analysis, evolution of emergency core cooling systems (ECCS) from active to fully passive systems to remove the decay heat, and development and consolidation of the best-estimate safety analysis methodology. With substantial increase in computing power, the computational fluid dynamics (CFD) tools for single-phase and multi-phase flows are being used more these days to address some of the important reactor thermal-hydraulics phenomena which could not be analyzed earlier using the traditional one-dimensional or coarse three-dimensional analysis tools. Development of multi-physics methodology encompassing neutronics, thermal-hydraulics, thermal-mechanical and coolant chemistry has also started. This front matter contains the series editors’ preface, contents, abstract, and acknowledgements. This monograph summarizes the major developments on nuclear reactor thermal-hydraulics over the last fifty years, primarily for the water-cooled reactors, and provides a direction for the future thermal-hydraulic developments for water-cooled, including small modular reactors or SMR, and Generation IV reactors. eBook keywords: Nuclear reactor, thermal hydraulics, cooling systems, computational fluid dynamics, modular reactors, neutronics, coolant chemistry, nuclear fission, heat generation, reactor thermal hydraulics. Nuclear power plants, which produce electricity with almost zero greenhouse gas or carbon emissions at stable and competitive costs, are operating in more than thirty countries throughout the world. According to International Atomic Energy Agency (IAEA) Reference Data Series No. 1 2015 Edition, in 2014, 438 nuclear power reactors with the total installed capacity of 376 GWe produced about 2410 TWh of electricity representing 11.1% of the total electricity produced in the world from all energy sources. Some countries in Europe (e.g., France, Slovakia, Hungary, Ukraine, Belgium, and Sweden) produced more than 40% of their electricity from nuclear. The United States of America (USA) produced about 800 TWh of electricity, representing 19.5% of total U.S. electricity production, from its 99 nuclear reactors. This represented over 60% of electricity produced from low carbon-emission sources including hydro, solar and wind. Percentage of electricity from nuclear is also increasing in Asia, particularly in China, Korea and India. Thus nuclear power constitutes an element of the solution to global warming and a means of delivering electricity to both developed and emerging countries. As discussed in Section 1.3.1, one of the desired features of an “ideal” reactor coolant would be “well-defined phase state” or no phase change during steady state or transient operation. Only gaseous coolants such as helium or carbon dioxide meet this criterion. Liquid metal cooled reactors such as Sodium-cooled fast reactors (SFR) and Molten salt reactors (MSR) are designed to avoid phase change or boiling in the reactor core so that the coolant remains as a liquid for all practical purposes. Determination of cladding temperature for these cases is simple once the surface heat flux, coolant bulk temperature and the local heat transfer coefficients are known (see Eq. 1-3). Well-established single-phase heat transfer correlations such as Dittus-Boelter correlation or those due to Gnielinski are used for gases and liquids (e.g., water) with Prandtl number on the order of unity. For liquid sodium with very low Prandtl number (on the order of 0.01 or lower) different heat transfer correlations based on experimental data are used. It was mentioned in Section 1.1 that all nuclear reactors generate “decay heat” even after “shutdown.” In order to remove this decay heat, the reactors are equipped with safety systems or Emergency Core Cooling Systems (ECCS). Determination of this decay heat as a ratio of the reactor power after shutdown to the initial or rated reactor power has undergone a progressive series of refinements through the publication of ANSI/ANS standards. Figure 3-1 shows the result derived from the 2005 ANS standards for a UO2 – fueled PWR with 4.2% enrichment after a long period of operation as fitted by Shwageraus and Fridman and as depicted in Figure 3-11 of Reference 112. Nuclear reactor safety analysis is the vehicle to determine if the safety systems discussed in Chapter 3 would perform satisfactorily under all Anticipated Operational Occurrences (AOOs) or transients and Design Basis Accidents (DBAs). Historically, the U.S.A. has been the leader in nuclear safety analysis as a number of nuclear power plants for commercial electricity generation started to operate in the U.S.A. in the 1960s. Two PWRs (Yankee-Rowe and Indian Point Unit 1) and one BWR (Dresden Unit 1) each of net electrical output of around 200 MWe started operation in early 1960s. Larger U.S. nuclear power plants of net electrical output of 500 MWe or more (e.g., Haddam Neck (PWR), Oyster Creek Unit 1 (BWR)) started operation in late 1960s, after which many large nuclear power plants (∼1000 MWe) were ordered, heralding the modern era of nuclear power plant operation using the Generation II reactors described earlier. At that time, the U.S. Atomic Energy Commission (USAEC) had the dual role of promoting and regulating the U.S. nuclear power plants. Nuclear power is a major low carbon-emission energy source for generation of electricity in the world, and it is expected to remain so in the foreseeable future. However, the nuclear power must be safe, reliable and economic compared to the other energy sources including hydro, wind, solar, natural gas, and even coal. This back matter contains the references, author biography, and index

Power uprate and life extension are ongoing concerns for nuclear power stations due to strengthened safety and regulatory requirements. While great successes have been achieved, huge costs have been reported and numerous technically challenging questions have been raised. With regard to piping, most issues concern whether a conducted finite element analysis is sufficiently good and whether relevant code requirements are reasonably met. This monograph addresses a few issues essential for achieving a consistent Design-By-Analysis in accordance with the ASME Boiler & Pressure Vessel Code: nonlinear alternative rules, fatigue, thermal ratcheting, dynamic loads, strain-based criteria, nonlinear finite element procedures, supports under combined stresses, and others. Throughout the text, the authors endeavor to seek more reliable and consistent alternatives using advanced non-linear finite element analysis. Unlike other texts on piping, this monograph is focused on commercial software and current practices for power uprate and life extension of aging nuclear power facilities. This front matter contains the table of contents, acknowledgments, preface, series editors’ preface, and nomenclature The structural integrity of pipe structures is of great concern in various industries, and particularly in the nuclear power industry. A pipe structure may include thousands of components, but the failure of just one of those components has the potential to shut down an entire nuclear power station or even to cause a severe threat to public safety. Although pipe structures are critically important, piping analysis has been generally considered as an area of low technology with limited academic relevance. In accordance with ASME III, a nuclear piping system which is found to be disqualified, i.e., overstressed, in design evaluation using linear analysis software can still be qualified through using a set of alternative rules. The application of the alternative rules involves generally nonlinear finite element analyses where nonlinear conditions such as plasticity are taken into account. This chapter and subsequent chapters will clarify these alternative rules. In particular, we will address unclear or inconsistent issues in the Code, and will describe nonlinear finite element procedures which are required when using the alternative rules. The general design evaluation rules for Class 2 piping are given in ASME III in articles NC-3600, and NC-3652 through NC-3655. The rules for Class 3 piping, given in articles ND-3600 of the Code, are basically the same as those for Class 2 piping (NC-3600). (See Slagis for an excellent review.) Stress limits are identical. We note minor differences regarding some requirements (for example, welding requirements), but significant differences regarding other requirements (for example, material quality requirements). For structural design evaluation for mechanical, thermal and other secondary loads, requirements and criteria for Class 2 piping are applicable for Class 3 piping. The nonlinear alternatives for structural design evaluation of Class 1 and 2/3 piping which were clarified and discussed in Chapters 2 and 3 generally involve nonlinear finite element analyses where nonlinear conditions (for example, plasticity) are taken into account. In the preceding chapters, we have attempted to clarify design requirements and nonlinear finite element analysis procedures by categorizing the design rules in ASME III into linear and nonlinear design rules. In particular, we have discussed nonlinear design rules. Our goal has been to overcome the difficulties that we have experienced when nonlinear design rules have been applied in conjunction to our power uprate and life extension tasks for aging BWR reactors in Sweden. Design rules and requirements for piping supports are given in Subarticle NF of ASME III. The rules and requirements are formulated in two sets by categorizing piping supports into two types: Linear Type and Plate and Shell Type . The categorization appears to be made with reference to the structural difference of supports, which was possibly the initial intention. In practice, in many cases the categorization is based on the approach of structural analysis which is employed. This back matter contains the references, index, and about the authors

In this book, expert authors recognize the renewed relevance of nuclear power in the U.S. after decades of stagnation. It was imperative to develop an up-to-date, scholarly work on containment structures, incorporating the underlying codes, regulations, safety significance, history, design philosophy, design experience, operating experience, and application to new design. This book will benefit the nuclear industry, as it transitions to a new generation of designers, constructors, and regulators. The book will be a valuable asset to the nuclear utilities, nuclear regulators, A/Es, and international organizations involved in the design and construction of Nuclear Power Plants (NPPs). This front matter contains the acknowledgments, contributor biographies, table of contents, and preface. This chapter describes the historical description of the evolution of nuclear reactors in the United States. A detailed historical narration is provided in Reference 1. The chapter also provides a brief description of the U.S. Atomic Energy Commission’s demonstration program, in which a number of reactor designs, other than the light water reactor designs, had been investigated, and for one reason or the other, eliminated from consideration. Chapter 2 of the book provides an in-depth historical background of regulations related to the licensing of nuclear power plants. At the time (circa 1952), only experience the AEC had was related to the small-scale government reactors and the demonstration reactors (see Table 1.1 of Chapter 1). With this knowledge and help from the research conducted at the national laboratories (e.g., Argon, Idaho), the AEC embarked on developing regulations that could apply to the commercial power reactors. One of the constraints imposed by the U.S. Congress was that the regulations should not be so restrictive that they would discourage the development of a private nuclear industry. At the same time, the regulations should be adequate to protect the public health and safety. During the period 1954 to 1962, the AEC’s review of the construction permit applications and development of regulations, were conducted side by side. By 1962, the AEC developed workable regulations that were used by the applicants of construction permit for nuclear power reactors. This chapter provides a summary of analysis and design and pre-operational inspection and testing requirements for concrete and steel containments. Both conventionally reinforced as well as prestressed concrete containments are included. Containments are intended to provide a leak-tight boundary against accident pressure loading which essentially dictates the generally used cylindrical shape. The size of the containment and available free volume depends upon equipment layout and operational requirements. The available free volume within the containment affects the magnitude of pressure that can be generated during an accident. In general, conventionally reinforced concrete and steel containments have been used for small containments with low to moderate internal design pressure. But large containments with relatively high internal design pressure require either a prestressed concrete or a steel containment. In general, selection of containment type (conventional concrete, prestressed concrete or steel) depends upon considerations for the equipment size and layout, design pressure and other loadings, construction cost and schedule. Containment reliability can be assured by periodic inspections and performance testing of containment components. This chapter describes the inspections and testing to be performed to assure that the containment will be capable of performing its intended function under the postulated loadings. The chapter also describes the deterministic inspection criteria and quasi risk-informed criteria so the uncertainties related to the use of risk-informed criteria are adequately addressed. As of August 2011, there were 104 commercial nuclear power reactors licensed to operate in 31 states in the United States. Initial operating licenses in the United States are granted for a period of 40 years. In order to help assure an adequate energy supply, the USNRC has established a timely license renewal process and clear requirements that are needed to ensure safe plant operation for an extended plant life. The principals of license renewal and the basic requirements that address license renewal are identified as well as additional sources of guidance that can be utilized as part of the license renewal process. Aging management program inspections and operating experience related to the concrete and steel containment structures are provided. Finally, several lessons learned are provided based on containment operating experience. Numerous studies have been conducted at Sandia National Laboratories and around the world to examine the severe accident behavior of containments through containment structural testing and modeling. Large-scale testing has been conducted on both prestressed concrete containments as well as on various steel shell designs. Since the containment includes the entire pressure boundary and not just the steel or concrete shell, some additional testing has been conducted on components (e.g., bellows, hatches). One of the goals of these test programs was to provide data for the calibration of computational models. Therefore, the accompanying computational analyses efforts for a number of the testing programs are also discussed. In addition, a summary of work examining the effects of containment degradation is provided. Various forms of degradation (e.g., liner corrosion) have been observed throughout the U.S. fleet of nuclear power plants. The effects of degradation on the severe accident response are examined in several studies using both deterministic and probabilistic analyses. These efforts describe the effect containment degradation has on the severe accident failure pressures, risks, and consequences. The Nuclear Regulatory Commission (NRC) defines three types of reactor accidents that have to be addressed to ensure public health and safety; (1) design basis accidents, (2) beyond design basis accidents and (3) severe accidents. A design basis accident is a postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures, and components necessary to ensure public health and safety. Loss of coolant accidents (LOCAs) are those postulated accidents that result from the loss of reactor coolant at a rate that is in excess of the capability of the reactor coolant makeup system. LOCAs are from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system. The containment structures are required to withstand the LOCA loads, as well as the loads imposed by the design basis natural phenomena, without exceeding conservatively calculated acceptance criteria. Containment structures are designed to remain functional when subjected to the loading from the natural events, such as earthquakes, flood water heights and associated effects, and high winds. Traditionally, the design bases for these natural hazards, particularly for seismic and flooding, have been evaluated using deterministic methods considering maximum historical events and accounting for uncertainties in characterizing such events. Our understandings of the natural hazard phenomena and our experiences from actual events have shown that the design bases are not bounding events and could be exceeded, albeit with low likelihoods. The evolution of probabilistic hazard methods has given us an ability to characterize hazard in terms of frequency of exceedance. These methods, in conjunction with the risk assessment methods, allow us to evaluate performance of a plant when subjected to design bases or beyond design bases natural events. The advanced reactors are often referred to as Generation III/III+ which are evolutionary or passive designs of pressurized water reactors (PWRs) and boiling water reactors (BWRs). In contrast to earlier generations of reactors, the advanced reactors rely on standardized designs which are simpler and more rugged and incorporate passive or inherent safety features, making them easier to operate and less vulnerable to operational upsets, therefore further reducing possibility of core melt accidents. The standardization process also facilitates expediting licensing, reducing capital cost, and shortening construction time. This back matter contains the appendix and index.

Nuclear Engineering and Technology for the 21st Century - Monograph Series Jovica Riznic, Series Editor With more than 75 years of combined working experience in the area of reactor physics and safety, the intention of the authors of this monograph is to provide a practical book on reactor physics, particularly for the safe operation of aged CANDU reactors, with minimal mathematics or equations. The book gives a glimpse of first principles and their engineering application in reactor physics, for those who are interested in or are working in the Canadian nuclear industry. The book is also ideal as a reference for physicists, operators, regulatory staff, and for those who need to interact with reactor physicists at CANDU sites, nuclear laboratories, institutes, universities, or engineering companies. This book assumes prior knowledge of nuclear physics offered at the secondary level. As very few equations appear in the monograph, it is not considered suitable for specialists whose focus is only on calculations or on the development of software on reactor physics. Such readers should refer to the books listed in the bibliography at the end of the monograph. This front matter contains the Series Editors’ Preface, Contents, List of Tables, List of Figures and Acronyms, Abbreviations and Preface are available for free viewing by clicking on the PDF below. eBook keywords: CANDU reactor, reactor physics, nuclear reactors, nuclear engineering, physics of fission, neutron kinetics, on-power refueling, reactor regulating system, regional overpower protection, neutron overpower protection, Xenon effects From the educational point of view, there are many textbooks on reactor physics used at various universities in the world. However, most of these textbooks focus either on application to Light Water Reactors (LWRs), or on the theory and mathematics, with a significant number of equations and computational schemes. Or else they were written more than 20, or even more than 60, years ago, and therefore they do not reflect the evolution of reactor concepts and engineering requirements since then. All those categories of books are either difficult to follow for non-physicists working in the nuclear industry, or else are of little value for those who are interested in special features of CANDU reactor physics. Nuclear fission is the splitting of a (large) nucleus, with the release of energy. The nuclei of some heavy elements, such as U-238, do exhibit spontaneous fission in nature. However, the rate of such fissions is extremely low. The half-life of uranium is longer than 100 million years, and most of its decay is by alpha emission, so spontaneous fission is not a practical source of energy. Spontaneous fission is not of much use to us as an energy source! A nuclear reactor is designed to achieve the very delicate balance between neutron “production” (release) in fission reactions and neutron loss by absorption and leakage. A given neutron will be “born” in a fission event and will then usually scatter about the reactor until it meets its eventual “death” either by being absorbed in some material or by leaking out of the reactor. A certain number of these neutrons will be absorbed by fissionable nuclei and induce further fissions, thereby leading to the birth of new fission neutrons, that is, to a new generation of neutrons. The ratio of the number of neutrons born in a fission-neutron generation to the number born in the previous generation is called the effective reactor multiplication factor, k eff . The k eff characterizes the balance or imbalance in the chain reaction. Alternatively, k eff can be defined by the ratio of production rate to loss rate of neutrons in the reactor. These definitions are given below: There are 2 concepts related to the “age” of fuel: irradiation (fluence) and fuel burnup. The fuel irradiation in a given fuel bundle, denoted ω, is defined as the time integral of the thermal flux in the fuel during its residence time in the core. Another term for irradiation is fluence. Irradiation is also known as the thermal-neutron exposure of the fuel. The units of irradiation are neutrons/cm 2 , or more conveniently, neutrons per kilobarn, n/kb. Since the cut-off of the thermal-energy range may be defined differently in different computer codes, the fuel irradiation may vary from computer code to computer code, and caution must therefore be exercised when comparing irradiation values using different codes. In documents, it has been more and more usual to report values of fuel burnup rather than fuel irradiation, as burnup does not suffer from differences in definition between codes. For CANDU reactors, the control of the long-term reactivity and of the power is carried out by on-power refuelling, while the control of the short-term reactivity and of power is done by the RRS. The RRS is part of the overall plant-control system that maintains the reactor power at a specified level, or, when required, manoeuvres the reactor power between specified setpoints. The reactor power setpoint can be entered by the operator (in the reactor-leading mode) or it can be calculated automatically by the Steam Generator (SG) pressure-control program (in the turbine-leading mode). The RRS consists of the following main components: There are independent, separate, and diverse ROP/NOP systems for the two SDS. Each ROP/NOP system consists of a number of flux detectors (see Section 5.4.3) which provide prompt measurements of neutron flux throughout the core. The detectors are mounted inside assemblies that penetrate the core, perpendicular to the fuel channels, vertically or horizontally. The system for SDS1 uses vertical detectors, the one for SDS2 uses horizontal detectors. Detectors are judiciously distributed to monitor the neutron flux throughout the core. In the CANDU 6, a total of 58 ROP detectors are used: 34 for SDS1 and 24 for SDS2. The number and location of detectors in the core are selected in an analysis whose objective is to ensure that as small a number of detectors as possible protect the reactor by tripping a SDS when local high powers threaten reactor safety from any flux shape that could arise in the operating reactor, while at the same time providing adequate margin-to-trip (MTT) to avoid possible restrictions on reactor operating power. Most fission products absorb neutrons to some extent and they accumulate slowly as the fuel burnup increases, hence decrease the long-term reactivity. The neutron-absorbing fission-product xenon-135 has particular operational importance. Its concentrations can change quickly in a power maneuvre, producing major changes in neutron absorption on a relatively short-time scale (minutes). A change in power for an operating reactor generally alters local parameters in the reactor such as the temperatures of the fuel, moderator, and coolant. A change in any of these local parameters causes a change in reactivity that, in turn, affects reactor operation (a feedback effect). Local parameters help to understand the feedback reactivity components related to the core evolution. For the CANDU reactor, the most important local parameters are the following: Source neutrons are essential for reactor restart after a long shutdown. The term “source neutrons” applied to a particular time interval refers to a steady supply of neutrons, constant over the time interval of interest. This supply must be independent of the current or very recent fission rate, which can vary over the time interval. Thus, source neutrons exclude prompt neutrons and even delayed neutrons which originate in the fuel (i.e., those born in the fuel itself). This exclusion does not apply to delayed photoneutrons, which come from fissions that have occurred a long time before, and whose numbers are quite constant over the current time interval (further discussion of this point below). The power referred to most frequently in reactor physics is neutron power. Neutron power is essentially the fission rate multiplied by the average prompt energy released and recovered per fission (see Section 2.1.2). It is also called “prompt” power, as it appears very quickly following fission. We cannot measure neutron power directly, but we do monitor the neutron flux with ion chambers located outside the calandria and in-core flux detectors. These neutronic signals are calibrated to the thermal-power measurement which allows neutron power to be derived. Reactor physics aims to understand accurately the reactivity and the distribution of all the reaction rates (most importantly of the power), and their rate of change in time, for any reactor configuration. To do this, the multiplication factor (or, equivalently, reactivity) and the neutron-flux distribution under various operating conditions and at different times need to be calculated repeatedly. Most of the other parameters of interest (such as neutron reaction rates, power, heat deposition, etc.) are derived from them. They are governed by the geometry, the material composition and the nuclear data (i.e., the neutron cross sections, their energy dependence, the energy spectra and the angular distributions of secondary particles, etc.). For radiation-shielding calculations, additional photon interactions and coupled neutron-photon interaction data are required. This back matter contains the References and Bibliography.

This frontmatter contains the Contents, Preface, Acknowledgements, About the Author, Introduction and is available for free viewing by clicking on the PDF below. eBook keywords: heat transfer, electromagnetic waves, electromagnetic radiation, thermal energy, thermal radiation, radiation energy, ultraviolet radiation, infrared radiation, total hemispherical emissivity, electromagnetic wave spectrum. We are lucky to have our Sun. It is a yellow dwarf star and at its surface, called the photosphere, it has an average temperature of 5800 K and it acts like an ideal thermal radiator. The Sun is at a mean distance of 1.496×10 11 meters away from Earth. Sun’s gravity holds our solar system together. Also, we are very lucky to have an ideal Earth’s surface and an atmosphere around her that balance out the incoming total radiation power with the outgoing radiation power for us to be able to live our lives on this beautiful planet in stable temperatures. In this chapter, we will deal with transient cooling of vertically hanged hot special steel plates by radiation and natural convection heat transfer. In engineering calculations of radiation heat transfer, a real surface emitting thermal radiation at a temperature T s has a total emissive power less than the total emissive power of a black surface at T s , namely less than σ× T 4 s . The ratio of a real surface’s emissive power to a black surface’s emissive power at the same T s is called the emissivity of a surface. Emissivity of a surface is a function of surface temperature, wavelength and direction of radiation emission. In engineering calculation, we deal with total hemispherical emissivity, ε Ts , which is the emissivity of a surface that is averaged over all wavelengths and hemispherical directions and which is only a function of surface temperature as given in Equation 2-1. In this chapter, we will deal with thermal radiation exchange between two black-bodies which are perfect electromagnetic radiation absorbers and emitters. In order to determine the net exchange of thermal radiation between two black surfaces, we only have to know the configuration factor, F A 1− A 2 , between these two surfaces and their surface temperatures, T A 1 and T A 2 , as shown in Equation 3-1. A1 is the surface area of body number one. σ is the Stefan-Boltzmann constant which is 5.67×10 -8 W/m 2 -K 4 . The configuration factor is defined as the fraction of thermal radiation that leaves a surface and intercepted by another surface. Configuration factors are sometimes called view factors or shape factors in literature, see Refs [13] and [19]. Derivation of configuration factors which is a geometric relationship between surfaces can be found extensively in literature, Ref [33]. In this chapter, we will investigate the effectiveness of a rectangular and vertical thermal radiation heater to a rectangular and vertical elemental surface that is located at the base, Z=0, of the heater as shown in Figure 4-1. The elemental surface can be located at different X and Y locations at the base of the thermal radiation heater depending upon the desired temperature of the elemental surface. The thermal radiation heater A 2 has a rectangular surface with a height, H, and a width, W, where W<<H. The vertical elemental surface dA 1 has a rectangular surface with a height, H and a width, W. In this chapter, we will investigate center of glass heat transfer in vertical double pane windows. For contributions to total heat transfer from other parts of a window system such as edges of glass and window frames, see Ref [1]. It is assumed that vertical double panes are parallel, air tight and with no moisture condensation. Also, solar radiation strikes the panes at near normal incidence. In the present analysis, it will be assumed that the outer and the inner surfaces of the glass layer has the same temperature and they are isothermal, since the glass layers are thin, i.e. about 0.003 m thickness, and the thermal conductivity through the glass layer, i.e. about 300 W/(m 2 -K), which is two orders of magnitude higher than the overall conductivity through a double pane window, i.e. about 3 W/(m 2 -K). Steady-state heat transfer conditions are considered and heat flow is perpendicular to glass pane surfaces and it is one-dimensional. Radiation heat transfer ignores gas absorption between panes. Convective heat transfer assumes high aspect ratio panes. With above assumptions, center of glass heat transfer for a vertical double pane window is illustrated in Figure 5-1. Approximately 35% of energy released from a nuclear explosion is in the form of thermal radiation. Surface of a nuclear explosion fireball emits large amounts of ultraviolet, visible and infrared rays in the first few seconds after a nuclear explosion. Approximately 50% of energy released from a nuclear explosion is in the form of a pressure wave called a shock wave or it is also called a blast wave for air bursts. This pressure wave develops immediately after the burst and moves outward from the growing fireball. Approximately 5% of energy released from a nuclear explosion is characterized as the initial nuclear radiation that is produced in the first minute of the explosion and which produces high energy neutrons and gamma rays with wavelengths less than 10 -4 μm. Approximately 10% of energy released from a nuclear explosion is characterized as the residual nuclear radiation that emits gamma rays and beta particles due to delayed radioactivity of explosion residues. About 95 % of total solar radiation reaching Earth’s surface comprises of visible (380 to 750 nanometers, nm wavelengths) and infrared (700 to 1 x 10 6 nm wavelengths) regions of the electromagnetic radiation spectrum. Remaining 5 % of total solar radiation reaching Earth’s surface comprises of ultraviolet (UV) (100 to 400 nm wavelengths) region of the electromagnetic radiation spectrum. The UV region of the electromagnetic radiation spectrum is divided into three bands, namely UVA band between 315 and 400 nm wavelengths, UVB band between 280 and 315 nm wavelengths and UVC band between 100 and 280 nm wavelengths. Excessive exposure to solar UV radiation can cause very serious health effects on our skin, i.e. skin cancer, to our eyes, i.e. cataracts, and to our immune system. Also, exposure to solar UV radiation has cumulative adverse effect on human health. Fluorescent lamps for general lighting come in different powers, sizes and phosphors to emit visible radiant energy. A plastic cover called diffuser that is placed over the lamp effectively eliminates all ultraviolet radiation from the lamp. However, if there is no cover over the lamp and the glass envelope of the lamp is defective and has metal impurities in it, the ultraviolet radiation (UVR) emission from the lamp can cause significant cumulative damage to human eyes and skin. Safety standards for lamps and lighting equipment specify that the effective UVR exposure, i.e. spectrally weighted, has to be below 0.001 W/m 2 . This level of human exposure to UVR has no photobiological hazard to skin and eyes. ICNIRP, see Ref. [27], standard limits the UVR exposure energy to human skin and eyes at 30 Joules/m 2 . With this standard UVR exposure energy limit, the maximum effective UVR exposure duration can be calculated using Equation 8-1. Heat transfer in a thermos bottle mainly involves radiation and to a much lesser extent conduction and convection. Dewar invented the thermos bottle at the beginning of the twentieth century which is simply a stainless steel inner bottle covered by another stainless steel outer bottle which are separated by vacuum. The outer wall of the inner bottle and the inner wall of the outer bottle are coated with silver in order to minimize radiation heat transfer between the fluid inside the thermos bottle and the outside environment. See Figure 9-1. Two silver coated walls facing each other reflect electromagnetic waves between themselves in the infrared region and minimize radiation heat transfer from the medium inside the inner bottle to the outside environment or vice versa. There is some heat loss, i.e. less than 2.58% of total heat loss in the present example, due to conduction and convection from the lid of the thermos bottle where a small portion of the inner surface of the stainless steel inner bottle contacts the lid. Painting a vehicle involves several layers such as electrocoating that is applied to the virgin metal body, primer, base coat, clear coat, sealants and structural adhesives. In order to achieve the best paint performance, each layer of paint has to be cured in a bake oven according to paint supplier’s bake temperature versus time specifications. Accurate control of bake temperature versus time specifications is critical to the performance of a coating both physically and cosmetically. During the baking process, complete cross linking of the paint has to be achieved. In the present example, let us investigate a paint curing bake oven where heat is supplied to a particular vehicle’s painted body by hot air flow, i.e. by laminar forced convection heat transfer and by infrared radiation heat transfer lamps. In 2020, World energy consumption is estimated to be about 167,786 TeraWh/year. Only 1.25% of this energy consumption, namely 2,093 TeraWh/year, will come from our Sun through photovoltaic (PV) panels. Projection of yearly World energy consumption to 2040 show not much increase from the 2020 consumption. However, by 2040, energy coming from our Sun increases eight-fold to 9.66% of World energy consumption to 16,747 TeraWh/year mostly replacing coal and oil. This bright future in solar radiation energy consumption makes everyone eager to learn about PV panel system design and optimization of solar radiation energy conversion into electrical energy. This optimization process starts with the knowledge of solar radiation energy available at a particular location and at a particular time on Earth. There are numerous ways to convert solar radiation to thermal energy in a fluid. In this chapter only glazed flat plate solar radiation collectors will be treated. The function of a solar radiation collector is to absorb solar radiation energy by means of a high absorptivity, i.e. greater than 0.95, surface and convert it into thermal energy. The thermal energy is then transferred from the absorber surface to flowing fluid by means of high thermal conductivity tubes. The collector is protected from convective losses to the environment with a thin film coated glass cover that is highly transparent to incoming short wavelength radiation, i.e. greater than 0.85 transmissivity, but prohibits long wavelength radiation leaving the collector due to its very low emissivity, i.e. less than 0.1. Insulating material is placed along sides and back of the collector in order to minimize convective heat losses to the environment. Humans are very sensitive to changes in environmental temperature and humidity because our internal body temperature has to be kept at a narrow temperature range of 37 ± 0.5 C. Hypothalamus is the central control organ in the brain that controls our internal body temperature. When our internal body temperature rises above 37.5 C, our blood flow is increased to our skin in order to increase the heat transfer from our skin to the environment. At elevated internal body temperature sweating also starts and cools our skin which increases the heat transfer from the body core to our skin. When our internal body temperature falls below 36.5 C, our blood flow is decreased to our skin in order to decrease the heat transfer from our skin to the environment. Also when our internal body temperature falls below 36.5 C, muscle shivering can start in order to increase the internal metabolic heat generation. Thermal energy balance between a human body and the environment is shown in Equation 13-1. Swimming pools use solar collectors or natural gas heaters to maintain comfortable temperatures that are desired, i.e. between 20 C to 30 C. UV stabilized transparent plastic or bubble type covers are used during night to minimize losses due to evaporation heat transfer and convection heat transfer from pool water’s surface. Since water in a pool has a very large thermal mass, water’s temperature changes slowly. Therefore in the present analysis, steady-state heat transfer conditions will be analyzed. Also monthly averaged solar radiation absorbed by water, relative humidity of air, latent heat of evaporation and diffusion coefficient of water in air will be used. Conduction heat transfer from pool walls to earth will be neglected, since this mode of heat transfer is less than 2% as compared to major heat transfer mechanisms, such as incident solar radiation, evaporation from surface of pool water, convection losses from surface of pool water and infrared radiation losses from surface of pool water. With above assumptions, net heat loss from surface of pool water in a month can be estimated using Equation 14-1. In this chapter, radiant heating of spherical balls with different radii will be analyzed. A spherical ball (1) will be placed inside at the center of a spherical oven (2) which has a temperature controlled inner surface for infrared radiant heating. The spherical ball is heated by radiation in a vacuum environment, namely there is no convection heat transfer at the surface of the spherical ball. See Figure 15-1. In this chapter, steel spherical balls at different radii that were heated to 400 C (673 K) by infrared radiation in a vacuum environment in Chapter 15 will be cooled down slowly by natural convection and by infrared radiation heat transfer in unsteady state down to a temperature of 50 C (323 K). A hot spherical ball will be placed at the center of a large air chamber in which chamber air and chamber walls are kept at a constant temperature at 25 C (298 K). This kind of slow cooling down of hot metals are done in order to relieve stresses that are generated in the metal structures during their processing. In this chapter, we will investigate baking of a pizza by infrared radiation heat transfer from hot ceiling bricks of a hemispherical dome of a wood fired oven and by conduction heat transfer from hot bricks covering the floor which are contacting the bottom surface of a pizza crust. During the baking process, pizza will also lose heat by mass transfer, namely by evaporation, from pizza’s top surface to oven’s air. All heat transfer mechanisms acting on the pizza are in unsteady state from the moment the pizza is placed on the floor of the oven. Objects of baking a good pizza is not to burn the bottom of the crust and have a crispy bottom and also cook the dough and toppings to an exact water content at the end of baking process. We will assume the pizza to be a large disk, i.e. a diameter of D = 0.25 m, with a small thickness, i.e. 0.01 m, and use the one dimensional unsteady state heat transfer equation presented in Equation 17-1. In this chapter, we will investigate microwave radiation in cooking ovens. Microwaves, in the electromagnetic spectrum, fall between the infrared radiation at the high frequency end, i.e. 300 GHz at 1 mm wavelength, radio waves at the low frequency end, i.e. 300 MHz at 1 m wavelength. Different bands of microwaves are used in different technologies ranging from garage door openers, point-to-point wireless communication links, satellite and spacecraft communications to advanced weather forecasting by providing global view of weather patterns and surface temperatures. Cosmic microwave background radiation fills the entire universe and it provides us knowledge about the beginnings of our universe, namely about the Big Bang. theory. We receive naturally occurring background electromagnetic radiation continuously. Naturally occurring radioactive materials are found in air, water and soil. We are exposed to cosmic and terrestrial radiation especially in high altitudes. Electromagnetic radiation sources are also human made such as nuclear power generation, medical uses for treatment and for diagnosis. In this chapter, I will review electromagnetic radiation exposures resulting from medical diagnosis. Liquid oxygen is used widely in industrial and medical fields and also in spacecraft rocket propellants. Liquid oxygen storage is less bulky and less costly than the equivalent capacity of high pressure gas oxygen storage. The freezing point of liquid oxygen in one atmosphere of pressure is −219 C (54 K). The boiling point of liquid oxygen in one atmosphere of pressure is −183 C (90 K). In this chapter, let us investigate the radiation heat transfer from a cryogenic spherical container for storing liquid oxygen. The spherical cryogenic container has a concentric outer spherical shell and facing spherical surfaces “1” and “2” are silver coated in order to minimize radiation heat losses from the liquid oxygen as shown in Figure 20-1. Storing liquid oxygen in a cryogenic spherical container with a concentric spherical outer shell was analyzed in Chapter 20. In this chapter, let us insert a concentric spherical radiation shield with low emissivity surfaces in between the cryogenic spherical container “1” and outer spherical shell “2” in order to minimize radiation heat losses from the liquid oxygen as shown in Figure 21-1. In this chapter, temperatures of different horizontal surfaces, i.e. angle between the plane of the surface and the horizontal is zero degree, are determined at 21 st of June, i.e. 172 nd day of the year, at 12:00 noon solar time of the day, i.e. solar hour angle is zero degree, namely angular displacement of the Sun east or west of the local meridian due to rotation of Earth on its axis is zero degree, and at four different locations. The locations are Anchorage, Alaska at 61.17 degrees latitude north, Seattle, Washington at 47.61 degrees latitude north, Sacramento, California at 38.58 degrees latitude north and San Diego, California at 32.73 degrees latitude north. All four locations are very close to the sea level, namely altitudes vary between 9 meters and 35 meters. So altitude effects are neglected in the following calculations. Cell phones and digital cellular networks can produce high enough non-ionizing radio frequency electromagnetic radiation energy during long exposures, i.e. more than 1 hour, that can increase temperature of brain tissues high enough to affect the human nervous system. International guidelines allow a localized maximum exposure which is specified as 10 W/kg specific absorption rate (SAR), averaged over a 10-gram tissue for non-ionizing radio frequency electromagnetic radiation energy. See Ref. [18]. Radio frequency radiation includes radio waves and microwaves. In the new fifth generation (5G) wireless technology for cell phones and for digital cellular networks, electromagnetic radiation waves operate between 30 and 300 GHz frequency which can produce high radiation energy levels and damage human tissues. A radar, an acronym for RA dio D etecting A nd R anging, transmits from its transmitter short pulses of electromagnetic waves at radio and microwave frequencies, i.e. 30 kHz to 100 GHz, which may be reflected by objects on their path. Reflected electromagnetic waves come back to the receiving portion of a radar system to provide information about the shape and speed of the reflecting object such as airplanes, boats, missiles, speeding cars, etc. For example, radars in the 5 to 35 MHz electromagnetic wave transmission range are used to map currents in coastal waters. 2 to 4 GHz electromagnetic wave pulses are used for air traffic controllers. 8 to 12 GHz electromagnetic wave pulses are used for marine targets and missile guidance. 100 GHz electromagnetic wave pulses are used for high resolution meteorological observations and imaging. Rain particles also scatter portion of the electromagnetic wave pulses back to the receiver. This information tells where it is raining and how much precipitation there is. Doppler effect on these electromagnetic wave pulses can also determine the wind speed and direction of a rain storm. This back matter contains the References and Index Electromagnetic waves generate radiation energy, and they play very significant roles in our lives. Electromagnetic waves are studied in almost every scientific field from astronomy, agriculture, chemistry, medicine to physics. This book focuses on heat transfer aspects of electromagnetic waves. There are twenty-four chapters in this book with solutions to heat transfer problems from electromagnetic waves’ radiation energy, with different uses and cases related to our lives. Each problem solution also investigates the sensitivity of critical independent variables to governing dependent variables. This frontmatter contains the Contents, Preface, Acknowledgements, About the Author, Introduction and is available for free viewing by clicking on the PDF below. eBook keywords: heat transfer, electromagnetic waves, electromagnetic radiation, thermal energy, thermal radiation, radiation energy, ultraviolet radiation, infrared radiation, total hemispherical emissivity, electromagnetic wave spectrum. We are lucky to have our Sun. It is a yellow dwarf star and at its surface, called the photosphere, it has an average temperature of 5800 K and it acts like an ideal thermal radiator. The Sun is at a mean distance of 1.496×10 11 meters away from Earth. Sun’s gravity holds our solar system together. Also, we are very lucky to have an ideal Earth’s surface and an atmosphere around her that balance out the incoming total radiation power with the outgoing radiation power for us to be able to live our lives on this beautiful planet in stable temperatures. In this chapter, we will deal with transient cooling of vertically hanged hot special steel plates by radiation and natural convection heat transfer. In engineering calculations of radiation heat transfer, a real surface emitting thermal radiation at a temperature T s has a total emissive power less than the total emissive power of a black surface at T s , namely less than σ× T 4 s . The ratio of a real surface’s emissive power to a black surface’s emissive power at the same T s is called the emissivity of a surface. Emissivity of a surface is a function of surface temperature, wavelength and direction of radiation emission. In engineering calculation, we deal with total hemispherical emissivity, ε Ts , which is the emissivity of a surface that is averaged over all wavelengths and hemispherical directions and which is only a function of surface temperature as given in Equation 2-1. In this chapter, we will deal with thermal radiation exchange between two black-bodies which are perfect electromagnetic radiation absorbers and emitters. In order to determine the net exchange of thermal radiation between two black surfaces, we only have to know the configuration factor, F A 1− A 2 , between these two surfaces and their surface temperatures, T A 1 and T A 2 , as shown in Equation 3-1. A1 is the surface area of body number one. σ is the Stefan-Boltzmann constant which is 5.67×10 -8 W/m 2 -K 4 . The configuration factor is defined as the fraction of thermal radiation that leaves a surface and intercepted by another surface. Configuration factors are sometimes called view factors or shape factors in literature, see Refs [13] and [19]. Derivation of configuration factors which is a geometric relationship between surfaces can be found extensively in literature, Ref [33]. In this chapter, we will investigate the effectiveness of a rectangular and vertical thermal radiation heater to a rectangular and vertical elemental surface that is located at the base, Z=0, of the heater as shown in Figure 4-1. The elemental surface can be located at different X and Y locations at the base of the thermal radiation heater depending upon the desired temperature of the elemental surface. The thermal radiation heater A 2 has a rectangular surface with a height, H, and a width, W, where W<<H. The vertical elemental surface dA 1 has a rectangular surface with a height, H and a width, W. In this chapter, we will investigate center of glass heat transfer in vertical double pane windows. For contributions to total heat transfer from other parts of a window system such as edges of glass and window frames, see Ref [1]. It is assumed that vertical double panes are parallel, air tight and with no moisture condensation. Also, solar radiation strikes the panes at near normal incidence. In the present analysis, it will be assumed that the outer and the inner surfaces of the glass layer has the same temperature and they are isothermal, since the glass layers are thin, i.e. about 0.003 m thickness, and the thermal conductivity through the glass layer, i.e. about 300 W/(m 2 -K), which is two orders of magnitude higher than the overall conductivity through a double pane window, i.e. about 3 W/(m 2 -K). Steady-state heat transfer conditions are considered and heat flow is perpendicular to glass pane surfaces and it is one-dimensional. Radiation heat transfer ignores gas absorption between panes. Convective heat transfer assumes high aspect ratio panes. With above assumptions, center of glass heat transfer for a vertical double pane window is illustrated in Figure 5-1. Approximately 35% of energy released from a nuclear explosion is in the form of thermal radiation. Surface of a nuclear explosion fireball emits large amounts of ultraviolet, visible and infrared rays in the first few seconds after a nuclear explosion. Approximately 50% of energy released from a nuclear explosion is in the form of a pressure wave called a shock wave or it is also called a blast wave for air bursts. This pressure wave develops immediately after the burst and moves outward from the growing fireball. Approximately 5% of energy released from a nuclear explosion is characterized as the initial nuclear radiation that is produced in the first minute of the explosion and which produces high energy neutrons and gamma rays with wavelengths less than 10 -4 μm. Approximately 10% of energy released from a nuclear explosion is characterized as the residual nuclear radiation that emits gamma rays and beta particles due to delayed radioactivity of explosion residues. About 95 % of total solar radiation reaching Earth’s surface comprises of visible (380 to 750 nanometers, nm wavelengths) and infrared (700 to 1 x 10 6 nm wavelengths) regions of the electromagnetic radiation spectrum. Remaining 5 % of total solar radiation reaching Earth’s surface comprises of ultraviolet (UV) (100 to 400 nm wavelengths) region of the electromagnetic radiation spectrum. The UV region of the electromagnetic radiation spectrum is divided into three bands, namely UVA band between 315 and 400 nm wavelengths, UVB band between 280 and 315 nm wavelengths and UVC band between 100 and 280 nm wavelengths. Excessive exposure to solar UV radiation can cause very serious health effects on our skin, i.e. skin cancer, to our eyes, i.e. cataracts, and to our immune system. Also, exposure to solar UV radiation has cumulative adverse effect on human health. Fluorescent lamps for general lighting come in different powers, sizes and phosphors to emit visible radiant energy. A plastic cover called diffuser that is placed over the lamp effectively eliminates all ultraviolet radiation from the lamp. However, if there is no cover over the lamp and the glass envelope of the lamp is defective and has metal impurities in it, the ultraviolet radiation (UVR) emission from the lamp can cause significant cumulative damage to human eyes and skin. Safety standards for lamps and lighting equipment specify that the effective UVR exposure, i.e. spectrally weighted, has to be below 0.001 W/m 2 . This level of human exposure to UVR has no photobiological hazard to skin and eyes. ICNIRP, see Ref. [27], standard limits the UVR exposure energy to human skin and eyes at 30 Joules/m 2 . With this standard UVR exposure energy limit, the maximum effective UVR exposure duration can be calculated using Equation 8-1. Heat transfer in a thermos bottle mainly involves radiation and to a much lesser extent conduction and convection. Dewar invented the thermos bottle at the beginning of the twentieth century which is simply a stainless steel inner bottle covered by another stainless steel outer bottle which are separated by vacuum. The outer wall of the inner bottle and the inner wall of the outer bottle are coated with silver in order to minimize radiation heat transfer between the fluid inside the thermos bottle and the outside environment. See Figure 9-1. Two silver coated walls facing each other reflect electromagnetic waves between themselves in the infrared region and minimize radiation heat transfer from the medium inside the inner bottle to the outside environment or vice versa. There is some heat loss, i.e. less than 2.58% of total heat loss in the present example, due to conduction and convection from the lid of the thermos bottle where a small portion of the inner surface of the stainless steel inner bottle contacts the lid. Painting a vehicle involves several layers such as electrocoating that is applied to the virgin metal body, primer, base coat, clear coat, sealants and structural adhesives. In order to achieve the best paint performance, each layer of paint has to be cured in a bake oven according to paint supplier’s bake temperature versus time specifications. Accurate control of bake temperature versus time specifications is critical to the performance of a coating both physically and cosmetically. During the baking process, complete cross linking of the paint has to be achieved. In the present example, let us investigate a paint curing bake oven where heat is supplied to a particular vehicle’s painted body by hot air flow, i.e. by laminar forced convection heat transfer and by infrared radiation heat transfer lamps. In 2020, World energy consumption is estimated to be about 167,786 TeraWh/year. Only 1.25% of this energy consumption, namely 2,093 TeraWh/year, will come from our Sun through photovoltaic (PV) panels. Projection of yearly World energy consumption to 2040 show not much increase from the 2020 consumption. However, by 2040, energy coming from our Sun increases eight-fold to 9.66% of World energy consumption to 16,747 TeraWh/year mostly replacing coal and oil. This bright future in solar radiation energy consumption makes everyone eager to learn about PV panel system design and optimization of solar radiation energy conversion into electrical energy. This optimization process starts with the knowledge of solar radiation energy available at a particular location and at a particular time on Earth. There are numerous ways to convert solar radiation to thermal energy in a fluid. In this chapter only glazed flat plate solar radiation collectors will be treated. The function of a solar radiation collector is to absorb solar radiation energy by means of a high absorptivity, i.e. greater than 0.95, surface and convert it into thermal energy. The thermal energy is then transferred from the absorber surface to flowing fluid by means of high thermal conductivity tubes. The collector is protected from convective losses to the environment with a thin film coated glass cover that is highly transparent to incoming short wavelength radiation, i.e. greater than 0.85 transmissivity, but prohibits long wavelength radiation leaving the collector due to its very low emissivity, i.e. less than 0.1. Insulating material is placed along sides and back of the collector in order to minimize convective heat losses to the environment. Humans are very sensitive to changes in environmental temperature and humidity because our internal body temperature has to be kept at a narrow temperature range of 37 ± 0.5 C. Hypothalamus is the central control organ in the brain that controls our internal body temperature. When our internal body temperature rises above 37.5 C, our blood flow is increased to our skin in order to increase the heat transfer from our skin to the environment. At elevated internal body temperature sweating also starts and cools our skin which increases the heat transfer from the body core to our skin. When our internal body temperature falls below 36.5 C, our blood flow is decreased to our skin in order to decrease the heat transfer from our skin to the environment. Also when our internal body temperature falls below 36.5 C, muscle shivering can start in order to increase the internal metabolic heat generation. Thermal energy balance between a human body and the environment is shown in Equation 13-1. Swimming pools use solar collectors or natural gas heaters to maintain comfortable temperatures that are desired, i.e. between 20 C to 30 C. UV stabilized transparent plastic or bubble type covers are used during night to minimize losses due to evaporation heat transfer and convection heat transfer from pool water’s surface. Since water in a pool has a very large thermal mass, water’s temperature changes slowly. Therefore in the present analysis, steady-state heat transfer conditions will be analyzed. Also monthly averaged solar radiation absorbed by water, relative humidity of air, latent heat of evaporation and diffusion coefficient of water in air will be used. Conduction heat transfer from pool walls to earth will be neglected, since this mode of heat transfer is less than 2% as compared to major heat transfer mechanisms, such as incident solar radiation, evaporation from surface of pool water, convection losses from surface of pool water and infrared radiation losses from surface of pool water. With above assumptions, net heat loss from surface of pool water in a month can be estimated using Equation 14-1. In this chapter, radiant heating of spherical balls with different radii will be analyzed. A spherical ball (1) will be placed inside at the center of a spherical oven (2) which has a temperature controlled inner surface for infrared radiant heating. The spherical ball is heated by radiation in a vacuum environment, namely there is no convection heat transfer at the surface of the spherical ball. See Figure 15-1. In this chapter, steel spherical balls at different radii that were heated to 400 C (673 K) by infrared radiation in a vacuum environment in Chapter 15 will be cooled down slowly by natural convection and by infrared radiation heat transfer in unsteady state down to a temperature of 50 C (323 K). A hot spherical ball will be placed at the center of a large air chamber in which chamber air and chamber walls are kept at a constant temperature at 25 C (298 K). This kind of slow cooling down of hot metals are done in order to relieve stresses that are generated in the metal structures during their processing. In this chapter, we will investigate baking of a pizza by infrared radiation heat transfer from hot ceiling bricks of a hemispherical dome of a wood fired oven and by conduction heat transfer from hot bricks covering the floor which are contacting the bottom surface of a pizza crust. During the baking process, pizza will also lose heat by mass transfer, namely by evaporation, from pizza’s top surface to oven’s air. All heat transfer mechanisms acting on the pizza are in unsteady state from the moment the pizza is placed on the floor of the oven. Objects of baking a good pizza is not to burn the bottom of the crust and have a crispy bottom and also cook the dough and toppings to an exact water content at the end of baking process. We will assume the pizza to be a large disk, i.e. a diameter of D = 0.25 m, with a small thickness, i.e. 0.01 m, and use the one dimensional unsteady state heat transfer equation presented in Equation 17-1. In this chapter, we will investigate microwave radiation in cooking ovens. Microwaves, in the electromagnetic spectrum, fall between the infrared radiation at the high frequency end, i.e. 300 GHz at 1 mm wavelength, radio waves at the low frequency end, i.e. 300 MHz at 1 m wavelength. Different bands of microwaves are used in different technologies ranging from garage door openers, point-to-point wireless communication links, satellite and spacecraft communications to advanced weather forecasting by providing global view of weather patterns and surface temperatures. Cosmic microwave background radiation fills the entire universe and it provides us knowledge about the beginnings of our universe, namely about the Big Bang. theory. We receive naturally occurring background electromagnetic radiation continuously. Naturally occurring radioactive materials are found in air, water and soil. We are exposed to cosmic and terrestrial radiation especially in high altitudes. Electromagnetic radiation sources are also human made such as nuclear power generation, medical uses for treatment and for diagnosis. In this chapter, I will review electromagnetic radiation exposures resulting from medical diagnosis. Liquid oxygen is used widely in industrial and medical fields and also in spacecraft rocket propellants. Liquid oxygen storage is less bulky and less costly than the equivalent capacity of high pressure gas oxygen storage. The freezing point of liquid oxygen in one atmosphere of pressure is −219 C (54 K). The boiling point of liquid oxygen in one atmosphere of pressure is −183 C (90 K). In this chapter, let us investigate the radiation heat transfer from a cryogenic spherical container for storing liquid oxygen. The spherical cryogenic container has a concentric outer spherical shell and facing spherical surfaces “1” and “2” are silver coated in order to minimize radiation heat losses from the liquid oxygen as shown in Figure 20-1. Storing liquid oxygen in a cryogenic spherical container with a concentric spherical outer shell was analyzed in Chapter 20. In this chapter, let us insert a concentric spherical radiation shield with low emissivity surfaces in between the cryogenic spherical container “1” and outer spherical shell “2” in order to minimize radiation heat losses from the liquid oxygen as shown in Figure 21-1. In this chapter, temperatures of different horizontal surfaces, i.e. angle between the plane of the surface and the horizontal is zero degree, are determined at 21 st of June, i.e. 172 nd day of the year, at 12:00 noon solar time of the day, i.e. solar hour angle is zero degree, namely angular displacement of the Sun east or west of the local meridian due to rotation of Earth on its axis is zero degree, and at four different locations. The locations are Anchorage, Alaska at 61.17 degrees latitude north, Seattle, Washington at 47.61 degrees latitude north, Sacramento, California at 38.58 degrees latitude north and San Diego, California at 32.73 degrees latitude north. All four locations are very close to the sea level, namely altitudes vary between 9 meters and 35 meters. So altitude effects are neglected in the following calculations. Cell phones and digital cellular networks can produce high enough non-ionizing radio frequency electromagnetic radiation energy during long exposures, i.e. more than 1 hour, that can increase temperature of brain tissues high enough to affect the human nervous system. International guidelines allow a localized maximum exposure which is specified as 10 W/kg specific absorption rate (SAR), averaged over a 10-gram tissue for non-ionizing radio frequency electromagnetic radiation energy. See Ref. [18]. Radio frequency radiation includes radio waves and microwaves. In the new fifth generation (5G) wireless technology for cell phones and for digital cellular networks, electromagnetic radiation waves operate between 30 and 300 GHz frequency which can produce high radiation energy levels and damage human tissues. A radar, an acronym for RA dio D etecting A nd R anging, transmits from its transmitter short pulses of electromagnetic waves at radio and microwave frequencies, i.e. 30 kHz to 100 GHz, which may be reflected by objects on their path. Reflected electromagnetic waves come back to the receiving portion of a radar system to provide information about the shape and speed of the reflecting object such as airplanes, boats, missiles, speeding cars, etc. For example, radars in the 5 to 35 MHz electromagnetic wave transmission range are used to map currents in coastal waters. 2 to 4 GHz electromagnetic wave pulses are used for air traffic controllers. 8 to 12 GHz electromagnetic wave pulses are used for marine targets and missile guidance. 100 GHz electromagnetic wave pulses are used for high resolution meteorological observations and imaging. Rain particles also scatter portion of the electromagnetic wave pulses back to the receiver. This information tells where it is raining and how much precipitation there is. Doppler effect on these electromagnetic wave pulses can also determine the wind speed and direction of a rain storm. This back matter contains the References and Index

This chapter provides a commentary on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 3, 2019 Edition. It is intended to be used as a companion to Division 3 by Manufacturers and Users of high-pressure vessels, and will also provide guidance to Inspectors, materials suppliers, and others. The material is generally presented in the same order in which it appears in the Code. Comments are not given about each paragraph, but paragraph numbers are referenced as appropriate. The comments apply to the 2019 Edition. History The original author of this chapter was J. Robert Sims Jr. who updated the chapter for the second, third and fourth editions and was joined by Charles Becht V for the fifth edition. For this revision, Mr. Sims has been joined by S. Michael Cooch. Abstract There are five ASME codes that apply to the transportation of oil and gas by pipeline. ASME B31.4 covers the transportation of liquids such as crude oil and refined petroleum products. The standard also covers certain non-hydrocarbon liquids, namely anhydrous ammonia, carbon dioxide, and aqueous mineral slurries. This chapter discusses the 2019 Edition of B31.4. ASME B31.8 covers transmission and local distribution of natural gas. This chapter discusses the 2020 Edition of B31.8. ASME B31.8S is a supplement to B31.8 covering integrity management of natural gas transmission pipelines. This chapter discusses the 2020 Edition of B31.8S. B31G covers evaluation of the fitness for service of B31.4 and B31.8 pipelines affected by corrosion. This chapter discusses the 2020 Edition of B31G. B31Q covers the qualification of frontline personnel who work on B31.4 and B31.8 pipeline systems. This chapter discusses the 2021 Edition of B31Q. History This chapter was newly created for the fourth edition. Michael J. Rosenfeld is the original author and has updated all editions including the current online edition. Abstract This chapter covers ASME Boiler and Pressure Vessel Code Section VII, which presents guidelines for safe, reliable operation as well as avoiding unsafe conditions in power boilers. The guidelines apply to power boilers that produce steam for external use at a pressure exceeding 15 psig from the application of heat, which come from the combustion of fuels, from various hot waste gases, or from the application of electrical energy. The bulk of this chapter is devoted to the discussion on Subsection 1 of Section VII, which contains information about industrial, gas- or oil-fired, watertube boilers. Subsection 1 introduces the fundamentals of watertube boilers, the principles of boiler operation, the associated auxiliaries, and appurtenances required by ASME Code Section I. It also introduces the necessary instrumentation, the examinations to be performed by boiler operators and staff, and the requirements for repairs, alterations, and maintenance. Subsection 2 presents an overview of the control of chemical conditions within steam boilers. It provides information relating to internal cleaning and laying-up of boilers, internal corrosion due to deposits within boiler tubes, and corrosion cracking. The discussion covers the types of water treatment problems within boilers, and how each problem may be recognized and corrected so as to prevent boiler failures. Subsection 3 deals with firetube, electric, and utility boilers and Subsection 4 includes all information pertaining to coal-fired and other solid-fuel-fired boilers. Articles in Subsection 4 describe the types of ash and their removal and contain short generic lists of maintenance checks for these boilers. Subsection 5 is a glossary of the terms that may be used when referring to a power boiler and its associated facilities. Section VII is a practical boiler book, a good review for experienced personnel and a good teaching aid for less-experienced personnel. History The original author of this chapter was Edmund W. K. Chang. The second and third editions were updated by Edmund W. K. Chang and Geoffrey M. Halley. The fourth edition was updated by James T. Pillow. The fifth edition was completely revised by William L. Lowry. Barrie Mollitor updated this chapter for the current online edition. Abstract This chapter covers alternative rules to the construction of pressure vessels under Section VIII, Division 2. The Section is made up of nine parts and the organization within each part is as follows: rules and requirements, nomenclature, tables, figures, normative annexes, and informative annexes. Part 1 contains general requirements addressing the following subjects: reference standards, units of measurement, tolerances, and technical inquiries. Part 2 consolidates the responsibilities and duties of the User, Manufacturer, and Authorized Inspector. It also contains general requirements for report forms and maintenance of records. Part 3 contains all requirements related to materials of construction and is organized in a similar fashion to Part AM of VIII-2. The coverage includes materials permitted for construction of vessel parts, and supplemental requirements for ferrous materials, Cr–Mo steels, Q and T steels with enhanced tensile properties, nonferrous metals, and boltings and castings. The requirements of Part 4 provide design rules for commonly used pressure vessel shapes under pressure loading and, within specified limits, rules or guidance for treatment of other loadings. Design requirements for the application of the design-by-analysis methodology in VIII-2 are described in Part 5. Detailed design procedures utilizing the results from a stress analysis are provided to evaluate components for plastic collapse, local failure, buckling, and cyclic loading. Part 6 consolidates the various fabrication requirements from old VIII-2 and VIII-1. Part 7 contains the requirements for examination during construction of pressure vessels. Part 7A covers the responsibilities and duties for inspection and examination during construction of pressure vessels. Part 8, which contains the rules pertaining to the pressure testing of the vessel following completion of fabrication, describes requirements for hydrostatic testing and pneumatic testing. The requirements for overpressure protection of vessels built in accordance with this VIII-2 are provided in Part 9. Abstract 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. Abstract Section I of the ASME Boiler and Pressure Vessel Code provides rules for the construction of power boilers. This chapter provides an overview of the requirements of Section I and explains how they are applied and enforced. The discussion begins with an outline of the history, philosophy, and scope of Section I, how the ASME Code works, and how it is organized. The chapter then addresses the distinction between boiler proper piping and boiler external piping, and describes how and where Section I is enforced. This is followed by a discussion on the fundamentals of Section I construction. The fundamentals include permitted materials, design, fabrication, welding and postweld heat treatment, nondestructive examination, hydrostatic testing, third-party inspection, and certification by stamping and the use of data reports. The chapter ends by highlighting some of the items which may be included in a future publication of Section I. A few of these include changes to the requirements of pressure testing of replacement parts; changes to the requirements for marking of plate material used for shells, furnace sheets, and heads; and inclusion of corrosion allowance application criteria. History Martin D. Bernstein was the originator of this chapter for the first edition. Lloyd W. Yoder updated the chapter for the second edition. John R. MacKay updated the chapter for both the third and fourth editions. Ed Ortman and Jay Vattappilly updated this chapter for the fifth edition. Ed Ortman updated this chapter for the current online edition. Abstract This chapter presents various specifications for the materials used in the construction of boilers. The first section of this chapter traces the chronological evolution of materials and associated technologies, from the need for materials to accommodate riveted construction to the acceptance of fusion welding as a fabrication process. Included in this discussion are the application of advanced materials, the revisions to the basis for setting allowable stress values, and the acceptance of Material Specifications other than those approved by ASTM. The next two sections discuss Code Section II, Part A– Ferrous Material Specifications and Part B– Nonferrous Material Specifications . The discussion covers the following: Guideline on the approval of new materials under the ASME Boiler and Pressure Vessel (BPV) Code; Submittal of Technical Inquiries to the Boiler and Pressure Vessel Committee; and Guidelines on Multiple Marking of Materials. The fourth section provides the basis for the Specifications and Standards covered by Section II, Part C– Specification for Welding Rods, Electrodes, and Filler Metals , and their relations to the ANSI/AWS specifications. Included in the discussion are the electrode classification system, material descriptions, welding material applications, welding material procurement, and filler metal certification. The fifth section reviews important aspects of Section II, Part D– Properties . The discussion covers the properties of ferrous and nonferrous alloys used in the design of components for the BPV and Nuclear Construction Codes. The final section discusses Section II, Part A and Part B— International Material Specifications . ASME Section II includes rules for acceptance of material specifications of recognized National or International organizations other than ASTM. The section also discusses the process of adoption of the Committee for Standardization specifications in Europe with consequences on the corresponding ASME specifications. History Dennis Rahoi, Marvin L. Carpenter and Domenic A. Canonico were the original authors of this chapter. The second and third editions were updated by Marvin L. Carpenter, Elmar Upitis and Richard A. Moen. The fourth edition was updated by Elmar Upitis, Marvin L. Carpenter, John F. Grubb, Richard C. Sutherlin, Jeffrey Henry, C. W. Rowley and Anne Chaudoulet. The fifth edition was updated by Richard C. Sutherlin, Jeffrey Henry, and Anne Chaudoulet The current online edition was updated by Anne Chaudouet, Elmar Upitis and John Grubb. Abstract This chapter provides a commentary on the principles of the ASME Boiler and Pressure Vessel Code to help users apply common sense when using the Code. It provides engineers, owners, inspectors, regulatory institutions, and other Code users with a better understanding of the basic principles of the Code. The chapter begins with a discussion on the design factors used by the ASME Committee to determine allowable stress and the requirements for a Design Specification and a Stress Report. It then provides information on the seismic design requirements and describes the use of Service Levels A, B, C, and D to provide increased allowable stresses for upset, emergency, and faulted plant conditions, as appropriate. The chapter emphasizes the need for good engineering judgment by knowledgeable professional engineers. This is essential for understanding which computer output should be used or how the output must be modified. The chapter discusses the revision made in the Foreword to the ASME Code in the 1992 Addenda to specify that engineering dimensions require tolerances and that Code-specified dimensions are based on nominal sizes. A brief note is provided on the identification of the proper allowance for erosion and corrosion caused by service conditions. The chapter also points out concerns with the use of finite element analysis in design activities. Forming operations for pressure vessels and postweld heat treatment of some welded configurations are briefly covered. In addition, the chapter discusses the importance of performance-based quality assurance programs and the use of thermal relief devices to protect pressure vessels or piping against pressure that results from the heat of the fluid contained in the component. Other issues addressed include the evaluation of new hypothetical loads on equipment that has been Code Stamped, and the purpose of hydrostatic testing of pressurized equipment. Abstract Subsection NCA provides the general requirements for both Divisions 1 and 2 of Section III. It is applicable to all Classes of construction, covering steel vessels, storage tanks, piping, pumps, valves, core support structures, and concrete containment vessels. This chapter identifies how to classify components and how the jurisdictional boundaries of Section III define what is within and outside the scope of the Code. It discusses the use of Code Editions, Addenda, and Code Cases, and the requirements for Design Basis, Design and Construction Specifications, and Design Reports. The chapter addresses the responsibilities and Quality Assurance Program requirements of the different entities involved in constructing a nuclear power plant, from the manufacturer of materials to the Owner. Requirements for ASME accreditation, application of the ASME Code Symbol Stamp, and use of Code Data Reports are also described. The chapter then outlines the basis for exemptions, load combinations, responsibilities of the different entities involved in nuclear power plant construction, Certificate of Authorization Holders, and Quality System Certificate Holders. Finally, it 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. History The current online edition has been updated by Jason Lambin and reviewed by Kyle Blan. Abstract This chapter discusses the structural limits in Section III NB and Section VIII Division 2 as related to specific structural failure modes considered applicable to pressure vessel and piping design. The rules and requirements for Section III NB and Section VIII Division 2 are based on the design-by-analysis philosophy addressing specific structural failure modes of structural collapse, rupture, instability, fatigue, and progressive deformation. The chapter begins with a discussion on the interactions between materials, environments, mechanical, and thermal loads as aided by other disciplines since this is critical for every part of every pressure vessel design. It provides a brief history of development of Code analytic tools and discusses the use of elastic 3D finite element analysis (FEA) results and the application of elastic-plastic (EP)-FEA. The chapter then discusses the intent of the primary stress limits in the Code. Consideration is given to the definitions in the Code and the discussion in the Criteria document. The chapter then addresses why the primary-plus-secondary stress intensity range is necessary and shows how to apply it. Two procedures for calculating stress intensity are presented, and the impact of thermal stresses is discussed for definition and application. The chapter also discusses six areas in which the Code presents fundamental procedures and expands on these fundamentals by presenting guidelines for a singular view of acceptable procedures that are consistent with the original intent for Section III and consistent with the current rules. Finally, it comments on the elastic-plastic FEA methods to protect against plastic collapse, to determine protection against local failure, and to assess fatigue and ratchet limits. History Tom Ahl was the original author of this chapter for the first edition. Marcus N. Bressler updated this chapter for the second and third editions. Chakrapani Basavaraju updated this chapter for the fourth, fifth and sixth editions. The current online edition has been updated by Chakrapani Basavaraju and Jason Lambin. Abstract This chapter addresses unique features of pressure vessels and atmospheric and 0–15 psi flat bottom storage tanks, as presented in the ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, Subsections NC (Class 2) and ND (Class 3) Components. Emphasis is directed at discussing the generic design issues of flat bottom tanks. Subsections NC and ND are a combination of rules and requirements taken from Section III, Subsection NB and Section VIII. In addressing the Articles of Subsections NC and ND, the chapter presents comparisons, the most probable source of origin of the Code requirements, certain insights as well as contradictions that seem to exist, and the specific source document and some of the underlying theory. The chapter begins with a discussion on the material selection and qualification requirements as defined in NC-2000 AND ND-2000. It then describes the requirements for fabrication and installation, including forming, cutting, and aligning. In addition, the chapter outlines the requirements for nondestructive examination of NC for vessels and sidewall of tanks. It also discusses the requirements for hydrostatic testing and overpressure and vacuum protection of 0–15 psi tanks and atmospheric tanks. The rules governing nameplates, stamping, and data reports are then presented. Finally, the chapter provides a summary of changes made to the various editions of this Code. History John Hechmer was the original author of this chapter that was authored by Greg L. Hollinger for the second edition and David P. Jones for the third edition. The fourth edition is authored by David P. Jones and Chakrapani Basavaraju. Chakrapani Basavaraju edited the fifth and sixth editions. The current online edition has been updated by Chakrapani Basavaraju and Jason Lambin. Abstract This chapter describes the bases and provisions of the ASME Boiler and Pressure Vessel Code for Concrete Containments. It 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 chapter addresses reinforced-concrete containment behavior under internal pressure and temperatures of up to about 100°C. The discussion presented includes a comparison between prestressed and deformed-bar concrete reinforcements, the effects of elevated temperature on concrete, and the effect of radial and tangential shear in containment shells. It explains the analysis, design, and testing of concrete containment structures and provides information on the organization of the Concrete Containment Design Code contained in the Code Section III, Division 2, Subsection CC. The chapter summarizes the following requirements for concrete containments: general and specific construction or placement requirements for concrete, including the fabrication and placing requirements for steel reinforcement systems and the fabrication and welding requirements for liners; requirements for nondestructive examinations, procedure qualifications and evaluations; requirements for qualification and certification of nondestructive personnel; and general requirements for containment structural integrity test. The major areas to be considered in the future revisions of the Code are also highlighted in this chapter.on and certification of nondestructive personnel; and general requirements for containment structural integrity test. The major areas to be considered in the future revisions of the Code are also highlighted in this chapter. History John D. Stevenson was the original author of this chapter and updated it for the second edition. The third edition was updated by Joseph F. Artuso, Hansraj Ashar and Barry Scott. The fourth edition was updated by Arthur Eberhardt, Michael Hessheimer, Ola Jovall and Clayton Smith. The fifth edition was updated by Arthur C. Eberhardt, Clayton T. Smith, Ola Jovall and Christopher A. Jones. The current online edition was updated by Arthur C. Eberhardt and Christopher A. Jones. Abstract Quick-actuating closures have been used on ASME pressure vessels for more than 50 years. This chapter was written because of the number of lawsuits filed against manufacturers of quick-actuating closures even though these were operated by the owners of pressure vessels for 20, even 30 and more, years without any serious accidents. This is because, in order to bring forth litigation, lawyers would skillfully take words in the ASME Boiler & Pressure Vessel Code completely out of context. Or, some attorneys for the injured party may misinterpret the Code rules to accuse the manufacturer of not having complied with the Code when the closure was made. So, based on experience, the author warns the writers of the ASME Code to assure that the rules are clear, concise, and understandable to the common man. He hopes that the rules in Section VIII, Division 1 will incorporate many of the suggestions mentioned in this chapter. To understand the current Code requirements for quick-actuating closures, the author identifies each of the changes made to the Code rules in Section VIII, Division 1, from 1952 Edition to the 2007 Edition of the ASME Code. The chapter includes an Appendix that provides guidance in the form of recommendations for the installation, operation, and maintenance of quick-actuating closures. This guidance is primarily for the use of the Owner and the User. The safety of the quick-actuating closure is the responsibility of the user. This includes the requirement for the user to provide training for all operating personnel, follow safety procedures, periodically inspect the closure, provide scheduled maintenance, and have all necessary repairs made in a timely fashion. The chapter also contains guidance for use by the designer. The chapter concludes with a list of suggested future changes in the Code for quick-actuating closures on pressure vessels. History Roger Reedy is the original author of this chapter introduced in the third edition and has updated it for the fourth, fifth and the current online edition. Abstract This chapter presents the rules of the ASME Boiler and Pressure Vessels Code, Section IV, constituting the minimum requirements for the safe design, construction, installation, and inspection of low-pressure steam boilers and hot-water Boilers (which are directly fired with oil, gas, electricity, or other solid or liquid fuels). It is divided into five parts and two sub-parts including several appendices. Part HG addresses the material requirements, design, pressure-relieving devices, tests, inspection, stamping, instruments, fittings and controls, and the installation of low-pressure steam-heating boilers, hot-water-heating boilers, hot-water-supply boilers, and the boilers’ appurtenances. Part HF, dealing with requirements for heating boilers constructed of wrought materials, has two sub-parts: HW and HB. The requirements of HW are applicable to heating boilers and their parts fabricated by welding. HB addresses the requirements for steam-heating boilers, hot-water-heating boilers, hot-water-supply boilers, and the boilers’ fabrication by brazing. The requirements of Part HC apply to steam-heating boilers, hot water-heating boilers, and hot-water-supply boilers and their parts, which are fabricated primarily of cast iron. The requirements of Part HA apply to hot water boilers constructed primarily of cast aluminum. Part HLW addresses the requirements for water heaters in commercial or industrial settings for supplying potable water at pressures not exceeding 160 psig (1100 kPa) and temperatures not exceeding 210°F (99°C). For each part, general requirements, material specifications, design parameters and calculations, fabrication requirements, and information on testing, inspection, and stamping are given. Appendices contained in this chapter include methods of checking safety valve and safety relief valve capacity; examples of methods of calculating a welded ring reinforced furnace; examples of methods of computation of openings in boiler shells; glossary; and two examples of manufacturer’s data report forms. Abstract This chapter provides criteria and commentary for ASME Section VI, which presents guidelines for the safe and efficient operation of steam-heating boilers, hot-water-supply boilers, and hot-water-heating boilers after installation. These rules are nonmandatory unless they are adopted into the laws of a governmental jurisdiction. The chapter is divided into nine parts, along with the necessary figures and tables for each part. Part 1 covers the scope of the chapter, provides some background on the use of illustrations and manufacturer’s information, and includes a glossary of terms applicable to boilers, fuels, fuel-burning equipment, combustion, and water treatment. Part 2 deals with the classification of boilers based on their design, construction, and application. Different types of boilers, such as steel, cast iron, modular, and vacuum boilers, are described and illustrated. The requirements of accessories such as safety relief valves and pressure gages, and their installation are addressed in Part 3. Part 4 deals with the characteristics of common fuels used for combustion in boilers such as gas, oil, coal, and wood products. Part 5 provides information on the controls for managing the functions of the fuel-burning equipment, namely operating, limit, safety, and programming controls. Part 6 covers all the requirements of boiler-room facilities for safe operation and maintenance. Recommendations for the operation, maintenance, and repair of steam boilers are included in Part 7 in addition to the procedures for inspection of steam boilers. Part 8 deals with the operation, maintenance, and repair guidelines for hot-water boilers and tests and inspections of hot-water-supply and hot-water-heating boilers. Water treatment considerations, boiler water problems, the chemicals used for water treatment and their functions, treatment alternatives, and various procedures are discussed in Part 9. Abstract The rules for the construction of Section III are based on the design and construction rules in Section VIII of the ASME Boiler and Pressure Vessel Code. This chapter identifies that the ASME Section III Nuclear Code is completely compatible with Sections I and VIII. Before the Nuclear Code was published, Code Cases for Sections I and VIII were written for the first nuclear plants in the early 1950s. In 1959, the Code Case book addressed nuclear installations as follows: General Requirements for Nuclear Vessels 1270N; Safety Devices 1271N; Containment and Intermediate Containment Vessels 1272N; Nuclear Reactor Vessels and Primary Vessels 1273N; and Special Material Requirements 1274N. The explanations of these original Code Cases are given in the chapter. The chapter briefly discusses the three types of pressure vessels addressed in Section III, namely, N-131 Class A Vessels, N-132 Class B Vessels, and N-133 Class C Vessels, covering the aspects of design, materials, and inspection, among others. It presents a brief note on the changes made to Section III in the 1971 edition. The chapter also presents an appendix providing information on the five Code Cases in the form of inquiries and replies. It covers the original Commentaries for Section III and Section VIII, Division 2, which is based on Section III reactor pressure vessel rules with the same design factor 3. Abstract This chapter covers the ASME B31.1 Power Piping Code and provides additional insight into the proper use of the Code. The chapter intends to supplement and provide additional insight to the proper use of the Code. It begins with a discussion of the conditions or considerations for the design of piping systems; the basis of the allowable stresses, quality factors, etc. to be used for the design of piping systems; and the requirements for pressure design of piping components. The chapter briefly addresses various piping components or types of piping joints which might be limited to nonhazardous or nonflammable fluids, size limitation, pressure, and temperature limits. It also addresses the requirements for designing power piping systems subject to thermal expansion, contraction, or other displacement stress producing loads. The chapter then provides information on the factors to be considered during the design of pipe supports and those piping systems typically found in power generating facilities. It presents some examples of how the piping codes are interrelated for some common carbon steel and stainless steel Pressure Rating/Dimensional Standards, Material Forming Standards, and Material Grades. The chapter lists the acceptable standards for Code construction such as ASME, API, AWWA, ASTM, etc. It then addresses the requirements for fabrication, assembly, inspection, examination, testing, and maintenance of piping systems. A number of appendices are included in the Code. The Mandatory Appendices contain information on allowable stress, thermal expansion data, moduli of elasticity, stress intensification factors and rules for nonmetallic pipes to name a few. The Nonmandatory Appendices include information on safety valves, corrosion, and operation and maintenance and more. History Charles Becht IV was the author of “Chapter 16” titled “B31.1 Power Piping” for the original, second and third editions. Chapter 17 of the third edition was revised in its entirety and renumbered as Chapter 36 in the fourth edition. The fifth edition was authored by Jimmy E. Meyer and Joe Frey. The current online edition has been updated by Jimmy E. Meyer and Kristi A. Vilminot. Abstract This chapter provides commentary on a new division under Section III of the ASME Boiler and Pressure Vessel (BPV) Code. This Division 5 was first published as part of the 2011 Addenda to the 2010 Edition of the BPV Code. The chapter provides information on the scope and need for Division 5, the structure of Division 5, where the rules originated, the basis for the elevated temperature rules specified in Division 5, the various changes made in finalizing Division 5, and the future near-term and long-term expectations for Division 5 development. It reflects the 2015 Edition of Division 5. Division 5 identifies rules for high temperature reactors (HTRs) based on only two classifications, Class A for safety-related components and Class B for non-safety related but with special treatment components. It contains general requirements for both metals and graphite in Subsection HA, Subpart A and Subpart B, respectively. Rules for Class A metallic pressure boundary components, Class B metallic pressure boundary components, and Class A core support structures at both low temperature conditions (under Subpart A) and elevated temperature conditions (under Subpart B) are contained in Subsections HB, HC, and HG, respectively. Rules for Class A and B metallic supports are contained in Subsection NF, Subpart A. Finally, new rules for nonmetallic core support structures (graphite) are contained in Subsection HH, Subpart A. Consistent with current Code practice, the primary concern of Division 5 is the integrity of the components under design, operating conditions (including normal, upset, emergency, and faulted), and test conditions. Division 5 covers all construction aspects of these components, including administrative requirements, material selection and qualification, design, fabrication, examination, inspection, testing, quality assurance, and documentation. History Effective with the 2015 Edition of the ASME Boiler and Pressure Vessel Code, “Subsection NH of Div. 1 of Section III” was discontinued and the subject matter transferred to Subsection HB, “Class A Metallic Pressure Boundary Components,” Subpart B, “Elevated Temperature Service” of Div. 5 of Section III. Accordingly, the subject matter previously covered in Chapter 12 of Fourth Edition and prior editions is now covered in an expanded version of Chapter 17 “Division. The current online edition was updated by Robert I. Jetter and Mark Mitchell with contributions from the BPV Committee on Construction of Nuclear Facility Components (III). Abstract This chapter covers the qualification of welding, brazing, and plastic fusing procedures and the qualification of individuals performing these procedures as required by the Construction Codes of the ASME Boiler and Pressure Vessel and Piping Codes. It first traces the history of Section IX, outlines the organization of the Section, and reviews the scope of weld processes that Section IX addresses. Section IX contains rules for the qualification of the several weld processes such as oxyfuel welding, shielded metal arc welding, submerged arc welding, gas metal arc welding, gas tungsten arc welding, and plasma arc, among others. The chapter then describes the classification of ASME base materials as given in Section IX. It explains the general requirements for both the welding procedure and welder performance qualification as well as the variables applicable to the welding procedure and welder performance qualification. The chapter also discusses the requirements to perform Charpy V-notch impact testing of the weld procedure qualification. It outlines the requisites for conducting the following testing and examination methods for groove weld and pressure-retaining fillet weld procedure qualification: tension test, guided-bend test, visual examination, volumetric examination, macroexamination, and fracture test. The qualification rules for two special weld processes, corrosion-resistant weld metal overlay and hardfacing overlay, are also explained. General requirements for brazing procedure and brazer performance qualification and the variables applicable to brazing procedure and brazer performance qualification are subsequently addressed. Finally, the chapter summarizes the general requirements for fusing procedure and fusing machine operator performance qualifications and identifies the variables applicable to fusing procedure and fusing machine operator performance qualification. Abstract This chapter discusses the general requirements of Section XI of the ASME Boiler and Pressure Vessel Code that are applicable to all Code Classes of components, including steel vessels, piping, pumps, and valves, as well as concrete structures. It identifies applicability limits of Section XI, addresses interfaces with any applicable regulatory requirements, and explains the use of Code Editions and Addenda and Code Cases. The chapter describes periodic nondestructive examination (NDE) and pressure testing used to ensure the integrity of components (other than containment vessels) within the scope of the Code’s jurisdiction. The requirements of NDE are specified, ranging from personnel qualification to the overall conduct of the examinations, as are the types of NDE required, including sample expansion, and the frequency of examinations required, such as from the detection of defects. Periodic pressure testing and pressure tests following repair/replacement activities are addressed. The chapter also addresses the responsibilities and Quality Assurance Program requirements of the different entities involved in the examination and testing of a nuclear power plant. It analyzes controversial issues and topics of current concern, including the applicability of recent United States Nuclear Regulatory Commission Generic Letters and Information Notices and describes the ways in which the reader can use recent Section XI revisions to his or her advantage. The chapter references some published ASME interpretations to explain how the Code requirements can be applied to commonly encountered problems. It contains information on risk-informed inservice inspection and reliability integrity management programs for non-light-water reactors. The chapter presents a note on a new Section XI, Division 2, concerning high-temperature gas-cooled reactors, which is anticipated to be published in 2019. Abstract This chapter provides a summary of some of the significant requirements of the ASME Code Section III, Subsection NE Class MC vessels, and a commentary on these requirements. It begins with an overview of the scope of Subsection NE and the boundaries of its jurisdiction. The chapter then outlines the general material requirements for Class MC vessels that are specified in Article NE-2000. This is followed by a discussion on material toughness requirements and design and analysis requirements. The chapter provides information on Certified Material Test Reports and on certifications that must be performed by engineers who prepare design and overpressure protection reports. The chapter then describes the rules for the following analyses of containment vessels as defined in Subsection NE: fatigue analysis, buckling analysis, design by analysis, and plastic analysis. An outline of the rules for reinforcement of cone-to-cylinder junctions, overpressure protection, bolted flange connections, and welded connections, is also presented. The closing pages cover the following topics: tolerance requirements for some commonly used shells; fabrication requirements for each weld category as specified in NE-4240; welding procedure qualification; rules for marking, examining, and repairing welds; and nondestructive testing personnel qualification and certification. A number of Code Cases and references regarding the rules of Subsection NE are cited, with cross-references to other Code Sections and Subsections. History Kamran Mokhtarian was the original author of this chapter and revised it for the second edition. Roger Reedy updated it for the third through current online edition. Abstract 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. History John D. Stevenson authored the original chapter and updated the second edition. The third edition was entirely rewritten with a new title and updated by D. Keith Morton and D. Wayne Lewis who also updated the the fourth and fifth editions. The current online edition has been updated by Garrick J. Solovey and D. Wayne Lewis. Abstract 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. Abstract This chapter describes how the ASME Boiler and Pressure Vessel Code requirements are implemented for safety-related nuclear valves. It explains the relevancy of the ASME Code requirements for safety-related nuclear valves using the latest issue of the Code. The chapter discusses the role of system and component design engineers, as well as the integrity of the Manufacturer. It provides a historical perspective for the Code rules, cross-referencing other Subsections of the Code. The chapter states that the Owner’s responsibilities for system design play an important part in establishing the rules applicable to the Design Specification for each safety-related valve. It also covers the operational and qualification requirements for the procurement of these items from the Manufacturer. The chapter discusses these items for different service conditions with the aid of schematics and references. It then provides details of the ASME B16.34 historical development and provides the current reference dimensional standards from the 2013 and 2015 editions of the Code for use in the construction of “N” stamped valves based on the Code’s B16.34 reference. A list of widely used Valve Standards referenced in ASME B31 Piping Codes is also presented. The coverage on pressure-relief valves includes details on activity underway at Section III to clarify requirements related to pilot-operated and power-actuated pressure relief valves. It also includes details relevant to QME-1 relating to qualification testing required for active valves used in nuclear power plants. History Late Douglas B. Nickerson was the original author of this chapter. Marcus N. Bressler updated the second edition and third editions. Guy A. Jolly, updated this chapter specific to valves for 4th edition and 5th editions. The current online edition was updated by Jason Lambin and the ASME Working Group on Valves. Abstract This chapter provides a commentary on the criteria and basis for requirements of Subsections NB, NC, and ND for Piping. It presents information on the corresponding design processes and design specifications, with pertinent references, tables, and figures. The chapter addresses service limits and loading conditions classified as Levels A, B, C, and D. The Code rules assure that violation of the pressure boundary will not occur if the design specification satisfactorily addresses all issues necessary for Code compliance. This assurance is provided in the following ways: protection against catastrophic failure and protection against initiation and propagation of a crack or propagation of a Section III acceptable flaw through the pressure boundary. The chapter addresses how Section III provides this protection, any variations that exist or that have existed between the piping rules and the design-by-analysis rules, and what items the analyst should be concerned with in satisfying Code requirements. It helps alleviate the confusion regarding the proper application of ANSI B31.7 rules. As the requirements of NB-3600 are discussed, the important areas are compared to those of NB-3200. The chapter also shows that for primary stress protection, NB-3200 and NB-3600 are quite different, but in the secondary and peak stress areas, including fatigue, the basis for the rules is essentially the same although the analytical approach is quite different. It then covers rules for both Class 2 (NC-3600) and Class 3 (ND-3600) piping design. The chapter also explains the importance of a design specification. Finally, it discusses the need for a process of analysis with release of information that needs to be established to provide an orderly flow of information, to assure timely pipe fabrication and support delivery, and to issue stress reports at reasonable times. History Donald F. Landers was the original author who updated the chapter for the second and third editions. The fourth edition was revised by Jack R. Cole. The fifth edition was updated by Jack R. Cole and Mark A. Gray are. The current online edition was updated by Mark A. Gray and reviewed by George Z. Tokarski. Abstract This chapter provides users of the ASME Boiler and Pressure Vessel Code Section V, Nondestructive Examination (NDE), an insight into the significant Section V requirements, the NDE methods addressed therein, and the applicability of the ASME Code NDE methodology to Section V and other sections of the ASME book. It explains when and how to use the American Society for Testing Materials (ASTM) Standards. The chapter traces the history of Section V and outlines its organization. It highlights the need for users to understand the relationship of Section V to other ASME Code sections to ensure that all contract requirements are satisfied. Section V is divided into two basic subsections. Subsection A contains Articles 1 through 19, including Mandatory and Nonmandatory appendices that address general requirements, test methods, and specific Code requirements. Subsection B contains the ASTM Standards (Articles 22 through 31 with Mandatory Appendix II – Standard Units Used in Equations and Nonmandatory Appendix A, Guidance for the Use of U.S. Customary and SI Units in the ASME). The chapter covers each of the 19 articles, including the various NDE techniques such as radiographic examination, ultrasonic examination for welds and materials, liquid penetrant examination, magnetic particle examination, eddy current examination, visual examination, and leak testing. Additional NDE techniques discussed in the chapter are acoustic emission examination for fiber-reinforced plastic vessels and metallic vessels, acoustic emission monitoring of pressure boundary components, alternating current field measurement, magnetic flux leakage examination, remote field testing examination, acoustic pulse reflectometry examination, and guide wave examination for pipe testing. Eddy current array examination methodology is being considered for incorporation into Section V. Abstract This chapter discusses the current and future nonmetallic construction codes and standards using nonmetallic plastic polymers. It is divided into four parts. Part A provides an overview of the major nonmetallic materials used in structural applications, namely, thermoplastic materials, thermoset plastic materials, and graphite materials. It describes the importance of a design specification as well as its recommended contents and major aspects. Part A also provides information on two extensively utilized thermoplastic jointing techniques: butt fusion and electro-fusion. Part B and C cover NM-1 and NM-2 Standards, which prescribe requirements for the design, materials, fabrication, erection, inspections, examination, and testing of thermoplastic piping systems and glass-fiber-reinforced thermosetting-plastic piping systems, respectively. Included within the scope of the NM-1 Standard is thermoplastic piping which interconnects pieces or stages within a packaged equipment assembly. The contents and coverage of the NM-2 Standard address pipe and piping components that are produced as standard products, as well as custom products that are designed for a specific application. Fiberglass-reinforced plastic pipe and piping components manufactured by contact molding, centrifugal casting, filament winding, and other methods are covered. The organization of the NM-3 Standard, addressed in Part D, is modeled on the ASME Boiler and Pressure Vessel Code, Section II – Materials. ASME NM-3 addresses material specifications and material properties for both thermoplastic and thermoset plastic with glass reinforcing fibers. The NM-1 and NM-2 Standards reference NM-3 Standard for applicable material specifications and material physical and engineering property data. History Some of the material in this chapter had been originally covered in Chapter 3.6 in the Fourth Edition of the Companion Guide but consistent with the current activity of the ASME Code Committees it was considered

This book is primarily intended for design and development engineers. The emphasis of this book is therefore on information which is of practical use. For this reason, theories and methods which do not provide useable solutions are dealt only briefly though sufficient references are provided for more information about them. Effort is made to provide the best available information for the design of heat exchangers in a clear and concise manner. This information includes experimental data, theoretical solutions, and empirical correlations. Accuracy and range of applicability of formulas/correlations presented is stated. Clear recommendations are made for application of the methods presented. The proposed book is on heat transfer in two-phase systems. These include boiling, condensation, gas-liquid mixtures, and gas-solid mixtures. Two-phase heat transfer is involved in numerous applications. These include heat exchangers in refrigeration and air conditioning, conventional and nuclear power generation, solar power plants, aeronautics, chemical processes, petroleum industry, etc. In recent years, there has been increasing use of miniature heat exchangers for computers and other electronic intensive products. This front matter contains the Wiley-ASME Press series, contents and preface and is available for free viewing by clicking on the PDF below. eBook keywords: heat transfer, heat exchangers, two-phase systems, boiling, condensation, gas-liquid mixtures, gas-solid mixtures, tube bundles, heat pipes, cryogenic pipelines. The two-phase systems covered in this book include boiling, condensation, gas–liquid mixtures, and gas–solid mixtures. Condensation is the phenomenon of vapor changing to liquid. There are two basic modes of condensation, namely, film condensation and dropwise condensation. During film condensation, vapor condenses in the form of a continuous liquid film over the condensing surface. During dropwise condensation, vapor condenses in the form of drops covering parts of the surface. Heat transfer coefficients during dropwise condensation are much higher than those during film condensation. Hence dropwise condensation is desirable. However, most common fluids condense in the film mode over ordinary metallic surfaces. Most of the condensers used in the industry today operate with film condensation. The mechanism of film condensation is well understood, and the rate of heat transfer during film condensation can be predicted with reasonable accuracy in most cases. Much research has been and is being done to develop durable heat exchangers with dropwise condensers but has not yet achieved success. Pool boiling is boiling without any forced flow. A typical example is during heating of water in a pan placed over a stove. In pool boiling, any movement of liquid is by natural convection. It is therefore also known as natural convection boiling. Boiling of subcooled liquids can occur if the heating surface temperature exceeds the saturation temperature. Heat transfer coefficients in subcooled boiling are considerably higher than in single-phase convection. Many heat exchangers operate exclusively in the subcooled regime. A notable example is the steam generator of pressurized water nuclear reactor (PWR) for power generation. Newer PWR reactors also operate with subcooled boiling. More commonly, subcooled boiling occurs in part of the heat exchanger near the entrance and saturated boiling in the rest of the heat exchanger. This chapter is about boiling of saturated liquids prior to critical heat flux (CHF). As vapor quality increases at a constant heat flux, heat transfer coefficient generally increases until a limit is reached when heat transfer coefficient begins to decrease. This is the CHF point. Beyond this point, the surface is partially in contact with liquid and partly with vapor. As the thermal conductivity of vapor is much less than that of liquid, heat transfer deteriorates. When quality reaches 100%, the entire surface is covered by vapor and heat transfer coefficient is that for forced convection vapor flow. Critical heat flux (CHF) during pool boiling was discussed in Chapter 3. Similar phenomena also occur during forced convection boiling. A sharp deterioration in heat transfer occurs at some point with increasing heat flux or quality. The heat flux at which the deterioration in heat transfer occurs is known as the CHF and the corresponding quality is known as the critical quality. The mechanisms and prediction of CHF for various conditions and types of heat exchangers are discussed in this chapter. Included are flow inside channels such as tubes and annuli, coils, single tubes, tube bundles, plates, coils, falling films, jets, sprays, etc. The various regimes in pool boiling were discussed in Section 3.1 and shown in Figure 3.1.1. The regimes in flow boiling are basically the same. In a surface temperature-controlled system, critical heat flux (CHF) is followed by transition boiling in which heat flux decreases with increasing wall temperature. This continues until the minimum film boiling temperature (MFBT) is reached. With further increase in wall temperature, film boiling starts during which heat flux increases with increasing wall temperature. In a heat flux-controlled system, film boiling starts immediately after CHF, and thus there is no transition boiling. Heat transfer to mixtures of permanent gases with liquid flowing in tubes/pipes is frequently encountered in petroleum, nuclear power, chemical, food, and pharmaceutical industries. Injection of gas into liquid is also used for enhancing heat transfer. Hence prediction of heat transfer for such applications is of considerable practical importance. Topics discussed in this chapter include heat transfer during flow of pre-mixed gas–liquid mixtures inside channels as well as heat transfer with gas injected through the surface of tubes. Also discussed are cooling of objects of various shapes by mist cooling, cooling by jets, effects of gravity, and evaporation from water pools. Sections 8.2–8.6 discuss only non-metallic fluids. Liquid metal–gas mixtures are considered in Section 8.7. Fluidized beds are widely used in many commercial applications, such as catalytic reactions, drying, coating, combustion, power plants, and ore roasting. This back matter contains the Appendix and Index

This publication follows the phenomenal success of not only the four editions of the Companion Guide to the ASME Boiler & Pressure Vessel Code published by ASME Press, but also two related updated volumes. Thus, this is the third book that is also a “standalone-publication,” addressing Global Applications of the ASME B&PV Code. This book not only updates information of 16 chapters of the third volume of the third edition of the Companion Guide, but has additional 5 chapters selected for their unique features of ASME Boiler and Pressure Vessel Codes used internationally. This book has five parts addressing Global Applications of ASME B&PV Codes and Standards covering: Part 1: North America and Western Europe which includes Canada, France, UK, Belgium, Germany, Spain and Finland in addition to the Pressure Equipment Directive of the European Union Countries. Part 2: Central and Eastern Europe includes Russian, Czech and Slovakian Codes and Hungary. Part 3: South Africa. Part 4: Asia including Japan, Korea, Taiwan, India and China. Part 5: Special Topics is addressed by ASME Code experts to cover in four chapters: (i) Global Harmonization of Nuclear Codes and Standards; (ii) Global Flaw Modelling Characteristics; (iii) AREVA’s perspective of spent fuel storage in a “A Case Study of Dry Storage System for Used Nuclear Fuel; and finally in last chapter (iv) Has three parts in “Utilities’ perspective of spent fuel storage” – the first one is covers ENTERGY, the second part Pacific Gas and Electric (PG&E) and the last part has Ontario Hydro’s experiences. Thus different perspectives of the Spent Fuel Storage which are critical to the continuation of nuclear industry are addressed by various experts in this chapter. An index is provided at the end of this publication as a quick reference to topics occurring in this volume which will assist in reviewing the interconnection of the ASME Boilers & Pressure Vessel Code Sections / Subsections / Paragraphs to each country covered in addition to the special topics addressed in this book. In each chapter, discussions generally pertain to the current Code Editions of the country covered and latest ASME Codes and Standards unless noted otherwise by the chapter author(s). This front matter contains the dedication, acknowledgements, contributor biographies, contents, preface, and introduction. This publication follows the phenomenal success of not only the four editions of the Companion Guide to the ASME Boiler & Pressure Vessel Code published by ASME Press, but also two related updated volumes. Thus, this is the third book that is also a “standalone-publication,” addressing Global Applications of the ASME B&PV Code. This book not only updates information of 16 chapters of the third volume of the third edition of the Companion Guide, but has additional 5 chapters selected for their unique features of ASME Boiler and Pressure Vessel Codes used internationally. eBook Keywords: global, pressure vessel, boiler, codes, standards, nuclear, ASME BPVC, power, energy. Chapter 1 introduces the Canadian boiler and pressure vessel standards and discusses their relationship to the ASME Code. The chapter provides the reader with an outline of the Canadian standards and an appreciation of some of their important characteristics, without going into every detail of these publications. For further information, interested readers should consult the text of the applicable standard(s). Pressure equipment presents inherent hazards due to the potential energy of its content, especially for gases. Consequently most industrialized countries have, for many decades, regulated these products. In European countries, each national authority developed regulations that were significantly different from each other: some were very detailed (e.g., in France), some contained only general requirements (e.g., in the UK), and others included very detailed conformity assessment procedures (e.g., in Germany). This Chapter presents the basic philosophy of the French Codes listed below and explains how they are linked to the Pressure Equipment Directive (PED), which is mandatory in the European Union; a comparison with the ASME Boiler and Pressure Vessel Code is provided. The presentation of nuclear codes will be limited to those applicable to pressure equipment. All of the acronyms used are provided at the end of this Chapter. Present code PD 5500, formerly BS 5500 evolved partly from the well-known BS 1500 in the 1950s and BS 1515 first published in 1965; the latter permitted higher level allowable stresses and more advanced rules. In 1969, following a report from the Committee of Enquiry into the Pressure Vessel Industry, the British Standards Institution brought all the pressure vessel interests together under one general committee in order to rationalize the activity. This became PVE/ and presides over a large committee structure. There are a series of functional sub-committees who deal with specific aspects and a large number of technical committees as well as many additional sub committees and working groups. Most of these meet regularly. The technical committee PVE/1, Pressure Vessels, has overall responsibility for BS 5500. The functional committee PVE/1/15 Design Methods has an overall responsibility relating to ‘Design’ with particular reference to the design section of BS 5500 (Section 3). Belgium has been an important actor in the applications of nuclear energy since the beginning. Very early, the availability of uranium from the former Belgian colony has contributed to the birth in the US of both military and civil aspects of the same. After the Second World War, scientists and industrialists participated to the development of the peaceful aspects of this new industry; the Belgian utilities own and operate today seven nuclear power units (four in Doel and three in Tihange), which produce 55% of the electricity in the country. The Pressure Equipment Directive (PED) (97/23/EC) was adopted by the European Parliament and the European Council in May 1997. Initially come into force on 29 November 1999. From 29 May 2002 the pressure equipment directive was obligatory throughout the European Union. This Chapter presents the regulation of pressure equipment in Spain, in the non-nuclear industry as well as the activities in the codes and standards in the nuclear field. In both cases, emphasis is placed on periodic inspections and testing. There are four operating nuclear power plant units in Finland. The TVO power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213, PWR units at the Loviisa site. Seawater is used as the ultimate heat sink at both plants. All the units were commissioned between 1977 and 1982. A first known historical example of safety regulation activity in Russia took place in the beginning of 18th century. On December 10, 1719, Emperor Peter the Great founded a special Berg Board for regulation of booming mining business in Russia. All state and private business units were subjected to regulation procedures, including safety issues of manpower and taxes. Before that responsibility for mining with different requirements and procedures was on local authorities. In 1734 an outstanding Russian geographer and captain of artillery Vassily Tatishchev prepared first Mining Regulatory Guide that was used in regional districts up to 1804 when it was officially approved by the capital Saint-Petersburg City Administration. In 1804 Russian Empire was divided into five mining administrative districts with Berg Inspectors as chief district officers. Nuclear power plants (NPP) in Jaslovske Bohunice (440 MW) in Slovakia and Dukovany (440 MW) and in Temelín (1000 MW) in the Czech Republic (both in former Czechoslovakia) were built with an agreement between the former Czechoslovakia and USSR in context of mutual co-operation in building nuclear power plants. Within this agreement, understanding was to use Soviet design and production standards and rules, as far as it was possible, while following the entire NPP design. Thus, for this case only, Government Decisions from 1978, 1980 and 1985 defined these rules. Early in eighties, extended project of the technical standard documentation of nuclear power plants was organized by the International Economic Association “Interatomenergo” in Moscow. The association was set up to co-operate in the field of nuclear power between individual member states of the Council of Mutual Economical Co-operation (CMEA). After accident in the Chernobyl NPP, the whole project was managed by the “Gosatomenergonadzor” (GAEN) of the Soviet Union. Under control of the GAEN an entire complex of technical standard documentation ended in 1990, when GAEN finished the whole project at international level and consequently also in the USSR. Paks, Hungary’s sole Nuclear Power Plant (NPP) consisting of four Russian designed VVER-440 model 213 units, was commissioned in the mid eighties with a design life of 30 years. VVERs are Russian designed PWRs, 440 refers to the original nominal electric capacity in MW, model 213 means that it is the second generation of the VVER-440s. An overview of the codes and standards for pressurized equipment as used in the South African nuclear industry is presented. The applicable legislation, regulatory requirements, and the roles of the respective regulatory bodies governing the use of pressurized equipment will be discussed. Two pressurized water reactor (PWR) units constructed by Framatome (now Areva) are commercially in operation in the country. A historical perspective of the construction, licensing, and operational phases is given, with emphasis on the design and quality rules used and risk considerations allowed by the licensing framework. The nuclear and non-nuclear codes used will be presented and the approach to and deviations from the in-service inspection code requirements will be explained. Japanese boiler and pressure vessel (B&PV) codes and standards were issued as governmental regulatory rules based mainly on the ASME Boiler and Pressure Vessel Codes Section I, Section III, and Section VIII. The administrative aspects of the Japanese rules had been changed from those of the ASME Codes to meet Japanese national situation and social environment. Roles and functions for the regulatory codes are public safety, administrative property, and general descriptions. Ministry of Economy, Trade and Industry (METI) deals with power generation facilities, gas utility facilities, high pressure gas facility, etc. Ministry of Health, Labour and Welfare (MHLW) covers boilers, small pressure vessels, etc. This chapter describes the nuclear regulatory organizations, aspects of the regulatory authority including licensing, status of nuclear installations, Korean nuclear reactor regulatory framework and regulations referring to domestic and/or international industrial codes and standards in the area of pressure vessels and piping, and status of Korean Electric Power Industry Codes (KEPIC) that covers standards for design, manufacturing, operation, maintenance, and testing and inspection of nuclear and non-nuclear mechanical components, including pressure vessels and piping. A comparative assessment of the U.S. and Korean codes is also addressed. In addition, a comparison between the KEPIC code and its reference to codes of other foreign countries is provided. Energy and power needs in Taiwan are inadequate. Almost 99% of electric energy sources are heavily dependent on imports of oil, coal, and natural gas. The total power energy was 252.2 TW-hours in 2013, and the electricity generation was contributed by thermal power of 63.4%, nuclear power of 16.5%, co-generation of 15.9%, hydropower of 3.4%, and renewable energy of 0.8%. Within the thermal power of 63.4%, coal generated 34.4%, liquid natural gas 26.9%, and oil 2.1%. Contribution by coal power generation is large. Renewable generations of 0.8% consisted of wind and solar, which are small. However, the renewable generations, which were almost zero in 2000, increased drastically in 2013. Particularly, wind power has considerably developed. The Nuclear Power Program in India, at present, is based mainly on a series of Pressurized Heavy Water Reactors (PHWRs). Starting from Rajasthan Atomic Power Station, comprising of two units of 200 MWe Canadian-designed PHWRs in 1973, the program has come a long way with 17 PHWR units (which include two units of 540 MWe PHWRs) in operation and four units of 700 MWe PHWRs under construction. Narora Atomic Power Station commissioned in 1991 marked major indigenization and standardization of PHWR design. The current design plan includes 700 MWe capacity units. The choice of PHWRs in the current stage of India’s nuclear power program is based on long-term objectives to be achieved with the available uranium resources and industrial infrastructure. These reactors use natural uranium as fuel and heavy water as moderator and coolant. The nuclear power stations in India are planned as twin-unit modules, sharing common facilities such as service building, control building, turbine building, spent fuel storage bay, and so on. This chapter describes India’s transition from the current heavy and light water reactors to fast breeder reactors and then advanced reactors primarily based on thorium fuel that is available in abundance. Some details of the designs are provided. In China, nuclear power programs have been ramped up to a fast lane in recent years. Latest statistics suggest that 26 units are in operation with a total capacity of 23,060 MWe and 24 units are under construction with a total capacity of another 23,740 MWe. By 2020, the capacity is expected to be increased to about 80,000 MWe. Therefore, a set of nuclear power codes and standards is urgently needed. However, historically, ever since China developed its own 300 MWe technology, deployed in Qinshan I, it has introduced different technologies from various countries, namely, Russia (ASE-91 VVER), France (M310), Canada (CANDU6), and the latest from the United States (AP1000). Consequently, different international codes, such as American Society of Mechanical Engineers (ASME), Règles de Conception et de Construction des Ilôts Nucléaires REP (RCC), ГОсударственныйСТандарт CCCP (ΓOCT), Canadian Standards Association (CSA), were adopted. Table 17.1 depicts the application of the different codes in China. This chapter addresses the current status of the harmonization efforts underway among various global Standards Development Organizations (SDOs). At the time of this writing, the SDOs actively involved in this project include the American Society of Mechanical Engineers (ASME), AFCEN (representing Europe), the Japanese Society of Mechanical Engineers (JSME), the Korean Electric Power Industry Codes (KEPIC), the Canadian Standards Association (CSA), and ENES Nikiet (representing the Russian Federation). Flaws might arise during manufacture and/or operation of pressure vessel and piping components in thermal and nuclear power plants, petrochemical plants, pipelines, etc. If flaws are detected in these components, assessments are needed to determine the cause of flaw formation, the remaining life of the component containing the flaw, whether the flaws should be removed from the components for continued operation, or whether repair/replacement of the component is required. Flaw assessments are performed to not only demonstrate the structural integrity of the degraded components during operation but also to demonstrate quality assurance during production or flaw tolerance in the design phase. Consensus assessment rules on flaw evaluation are needed for industries, utilities, and regulators. These rules require consensus agreement by all stakeholders. Several consensus codes and standards or guidelines of fitness-for-service (FFS) rules for flaw evaluation have been published for inspectors, designers, engineers, fabricators, and regulators. When civilian nuclear power plants in the United States of America (USA) were initially constructed, the assumption was that the Federal government would ultimately take title to the used nuclear fuel (UNF), relieving nuclear power plant operators of the burden of managing UNF at their sites. As we all know, this has not been the case. Thus, the nuclear industry in the USA has been challenged with finding a solution for storing UNF on nuclear plant sites for long periods of time. According to NEI, there was approximately 74,260 metric tons of spent nuclear fuel being stored at various sites across the United States as of December 31, 2014. While the major technical considerations for a permanent repository for storage of spent fuel have been addressed, political concerns have blocked efforts to open such a facility within the United States. As a result, utilities operating nuclear power plants have to address continued storage of spent nuclear fuel for the foreseeable future. The aspects briefly covered in this part pertain to the role of Independent Spent Fuel Storage Installation (ISFSI) for spent fuel storage at (a) the decommissioned nuclear facility at Humboldt Bay Power Plant and (b) Diablo Canyon Power Plant. In Ontario Canada, the spent fuel storage system for power reactors is comprised of three phases. The first phase, which occurs immediately after discharge of the fuel from the reactor, is in the irradiated fuel bays (IFBs). The IFBs are a wet storage system in which the fuel is stored and cooled in large, light-water filled pools (Section 21C.3). The second phase is a dry storage system that is utilized after a minimum of ten years of decay time in the IFBs (Section 21C.4). In this phase, the fuel is stored in dry storage containers (DSCs), which are steel lined reinforced concrete containers. The final phase is the Deep Geologic Repository (DGR) phase and is intended to occur at the end of DSC phase (Section 21C.5). The DGR phase is intended to be the permanent stage for spent fuel storage. This back matter contains the index

This commentary discusses some of the considerations of the joint ACI-ASME Committee in developing the provisions of ACI Standard 359 and ASME B&PVC Section III, Division 2, Subsection CC, Article CC-3000 on nuclear construction in the 2013 version of the code. Emphasis is given to the explanation of provisions that may be unfamiliar to code users. Comments on specific provisions are made under the corresponding paragraph numbers of the code. The figures and appendices referred to in this commentary occur only in the commentary so that their numbering has no parallel in the code. Because the code is written and intended for use as a legal document, it does not present background details or suggestions for carrying out its requirements or intent. It is the intent of this commentary to at least partially fill this need. This commentary also directs attention to other documents that provide suggestions for carrying out the requirements and intent of the code. However, neither those documents nor this commentary are to be considered as a part of the code. This front matter contains the dedication, acknowledgments, contents, references and introduction. This commentary discusses some of the considerations of the joint ACI-ASME Committee in developing the provisions of ACI Standard 359 and ASME BPVC Section III, Division 2, Subsection CC, Article CC-3000 on nuclear construction in the 2013 version of the code. Emphasis is given to the explanation of provisions that may be unfamiliar to code users. Comments on specific provisions are made under the corresponding paragraph numbers of the code. The figures and appendices referred to in this commentary occur only in the commentary so that their numbering has no parallel in the code. Because the code is written and intended for use as a legal document, it does not present background details or suggestions for carrying out its requirements or intent. It is the intent of this commentary to at least partially fill this need. This commentary also directs attention to other documents that provide suggestions for carrying out the requirements and intent of the code. However, neither those documents nor this commentary are to be considered as a part of the code. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. The code presents criteria for the design of a safe, reliable nuclear power plant containment. These criteria have been extrapolated from conventional codes and experience that has been built up through involvement with actual containment designs and testing programs. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. The material presented herein summarizes the progress that has been made in the analysis, design, and testing of concrete structures. Many of the most outstanding papers presented from SMiRT 1 to SMiRT 9 are referenced and at times conclusions are summarized. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. The design examples are taken from projects carried out by Scanscot Technology AB during 1992-2008, except for the one shown in Figure 6 that is taken from the paper “Increased plastic strains in containment steel liners due to concrete cracking and discontinuities in the containment structure”, P. Anderson and O. Jovall, presented at 19th SMiRT Conference, August 2007. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. Significant radial shear in the containment shell occurs at the junction of cylindrical wall to base slab and the transition between hemispherical dome and wall or ring girder to dome and wall. The cylindrical wall at penetrations is also designed for radial shear due to pipe rupture forces and the reaction from design pressure and temperature. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. This design example is presented to show the methods employed and the factors considered to establish the level of prestress for a prestressed concrete containment. Calculations are also given which are performed to assure that structural integrity is maintained. It must be pointed out that this example does not analyze all possible material properties, or all load combinations and locations in the structure which must be investigated. Calculations here are samples only and individual containments should be thoroughly checked by the designer. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. This paper presents the historical development of tangential shear considerations for concrete containments in the United States. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. The information presented in the paper is based on the testing documented in items: (a) and (b) below. Tests were performed on concrete end anchorage regions for prestressing tendons with ultimate strengths of approximately 8900 kN. One test structure simulated a full scale concrete containment buttress and the other two test specimens were concrete blocks. The behavior of the test structure and specimens, when subjected to loading, was monitored by strain gages and dial gages. The testing illustrated that all of the amounts of reinforcing tested should be acceptable for the end anchor zones of large tendons presently used in prestressed concrete containment structures. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon. The report gives a thorough description of the liner plate together with figures illustrating all the applicable details. Problem areas and factors affecting the liner plate and anchorage system design are covered in great detail. The loads considered in the liner plate design are discussed and the approximate strains in the liner plate resulting from these loads are stated. eBook Keywords: nuclear, containment, structures, concrete, reinforced, steel, stresses, stress, loads, pressure, standards, codes, anchors, liners, forces, impact, tendon.

This comprehensive work written by ASME Codes & Standards experts was originally published as part of Volume 3 of the Companion Guide to the ASME Boiler & Pressure Vessel Code. This fully updated and expanded volume is now a stand-alone publication that addresses Continuing and Changing Priorities for the success of current and next generation Nuclear Reactors and Internals, License Renewal, Public Safety, and PRA Fuel issues. This book has four parts, the first dealing with continuing priorities of ASME Boiler and Pressure Vessel Codes and Standards and the remaining three parts dealing with changing priorities of the Code. The first part has 13 chapters written by 15 distinguished code authorities with updated information pertaining to continuing priorities. Distinctive in this part is the inclusion of new contributors who have added unique perspectives to the material that was first published in the third edition of the Companion Guide. Likewise the authors who previously updated the material in the third edition have contributed additional information. A new chapter on “seismic protection for pressure piping systems” reflects the importance of the topic especially in regard to aging nuclear reactors. Three distinctive parts of the book address changing priorities of ASME B&PV Codes and Standards. These are: “Changing Priorities;” “Lessons Learned;” and “Future Discussions.”In Part 2, “Changing Priorities,” there are nine chapters by 11 contributors who are not only Code experts but also recognized authorities in the topics covered by them. Part 3, “Perspectives of Lessons Learned,” has eight chapters by 11 authoritative contributors who are also distinguished in the topics addressed by them. In light of the aging nuclear reactors in operation in the U.S. and around the world and the need for new reactors, it is appropriate to synthesize the expert experiences relating to selected topics. Part 4, “Future Discussions,” is the culmination of the current thinking process for the use of nuclear energy for power generation. This part has two chapters written by five authors with recognized expertise in the subject matter. A unique feature of this publication is the inclusion of all author biographies and an introduction that synthesizes every chapter, along with an alphabetical listing of indexed terms. This front matter contains the dedication, acknowledgements, contributor biographies, contents, preface and introduction. This chapter provides an overview of the history of the ASME Boiler and Pressure Vessel (B&PV) Code from its inception through its incorporation of nuclear components and up to the present. Applicable rules for certification and accreditation were developed in parallel with the growth of the Code. From earlier coverage of boilers and pressure vessels, the nuclear initiative required coverage for piping, pumps, valves, storage tanks, vessel internals, and both component and pipe supports. On July 10, 1991, a Maintenance Rule, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants , was published by the U.S. Nuclear Regulatory Commission (the USNRC, or simply the NRC) in the Federal Register (56FR31324) as 10CFR50.65. This Maintenance Rule was developed to provide a regulatory framework for the NRC for mechanisms used to evaluate the administrative effectiveness of maintenance programs applied by nuclear power plant licensees. The NRC's overall objective is to ensure that the structure, systems, and components important to the safety of nuclear power plants are maintained properly so that plant equipment will reliably perform its intended safety functions. The Maintenance Rule is characterized as performance-based, for it provides focus on results rather than on programmatic adequacy. In this chapter, background and requirements for testing, evaluating and qualifying vibration in power plants are addressed. This includes the testing and evaluation methods included in the ASME Operating and Maintenance OM-3 “Vibration Testing of Piping Systems”, the acceptable stress limits included in the ASME Boiler and Pressure Vessel Code and the applicable regulatory requirements for piping vibration. Examples are provided of various types of vibration problems experienced in power plants. Piping systems are often complex structures that include straight pipe and a variety of supports, elbows and tees. A typical piping system might include 50 or so components along with many intervening lengths of straight pipe. Each of the components is subjected to a complex set of loadings. The evaluation of any component by the detailed analysis methods prescribed in NB-3200 would be an onerous task. Fatigue is one of the most frequent causes of failure in pressure vessels and piping components. Fatigue strength is sensitive to design details such as stress raisers, and to a myriad of material and fabrication factors, including welding imperfections. Fatigue is also sensitive to often unforeseen operating conditions such as flow induced vibrations, high cycle thermal mixing, thermal striations, and environmental effects. What is somewhat surprising is the large number of fatigue failures which are directly related to poorly chosen design and fabrication details. The ASME Code, and other International Codes and Standards have not been successful in preventing the use of design and fabrication details that are inappropriate for cyclic service. The ASME Code was one of the first Codes and Standards to treat design for fatigue life explicitly. In past editions of the handbook, this chapter has primarily addressed issues as they pertained to Section III of the ASME B&PV Code. In particular, it has addressed issues concerning the seismic response of nuclear facilities. Currently, many other facilities covered by the ASME B&PV Code have had to address these same issues, often with little guidance from the appropriate Code section. Modern LNG terminals, for example, have had to undergo extensive seismic reviews. These facilities contain equipment covered by Section VIII and B31.3 of the Code. Neither of these Code section give any guidance to the designer concerning seismic analysis, other than to require that seismic loads be addressed. In this revision of the chapter, some of these issues will be addressed. In addition, a new section discussing the use of computer software for analysis has been included. This section will address some of the issues associated with different computer codes used for different parts of a plant. 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 presents a brief summary of force predictions from fluid phenomena and behavior, both of which are significant to the formulations and criteria employed in the ASME Boiler and Pressure Vessel (B&PV) Code. It focuses on the nature of fluid forces imposed on vessels and piping systems from the standpoint of designers, who need guidelines for both the design and operational recommendations of fluid transport systems. Although flanged joints are essential components in nearly all pressurized systems, they are also one of the most complex. Many factors are involved in determining the successful design and operation of a bolted flange joint in service—namely, the interaction between the bolting, flange, and gasket, as well as important nonlinear variables such as friction and gasket properties. Safety considerations are, of course, extremely important; just as the Pressure Vessel and Piping Codes were developed with safety in mind, so, too, were the rules for bolted flange joints. The Code Bolted-Flange Joint Design Rules provide a method for sizing the flange and bolts to be structurally adequate for the specified design conditions; however, the rules do not address assembly and special requirements, nor do they guarantee a leak-tight joint for all transient operating conditions. The latter issue is discussed further in the sections on tightness-based design and flange-joint assembly. The discussion in this Chapter is a highly condensed version of the information presented previously in Chapter 52 of the 2 nd Edition of the Companion Guide to the ASME Boiler & Pressure Vessel Code [1]. The full text of the previous Chapter 52, i.e., Development of U.S. Regulations for the Transportation of Radioactive Materials — A Look Back over the Past 40 Year s, could not be reproduced here. Therefore, this Chapter offers a high-level overview of the information presented previously, including all of the appropriate references. Pipelines provide a safe and economic means of transporting fluids such as natural gas, crude oil, refined products, and natural gas liquids products across great distances and over all manner of terrain. As man-made systems, however, they can and do fail from time to time, sometimes with catastrophic effect. Figure 11.1 illustrates the aftermath of a recent gas pipeline rupture followed by a fire which resulted in the death of 12 people [1]. Such fatal events, however, are rare, accounting for less than 0.02% of the 45,000 transport-related deaths that occur annually in the United States [2]. The consequences of failures, when they do occur, are more commonly measured in terms of lost product and its effect if any, on the environment, as well as a disruption to business while a repair is made [3]. Repairs and environmental remediation, especially in sensitive areas, can however be very expensive. For example, the clean up cost of a recent relatively modest crude oil pipeline break in Canada amounted to $24 million, in addition to a $1.4 million fine levied on the company for causing environmental pollution. In recent years several high profile pipeline leaks and ruptures particularly on liquid hydrocarbon lines have created a poor public image of pipelines. On July 26, 2010 a 30 inch diameter pipeline conveying diluted bitumen ruptured near Marshall in Michigan spilling almost 20,000 barrels into Talmadge Creek, which flows into the Kalamazoo River. Fortunately the response effort to this, the largest onshore spill in US history prevented bitumen reaching Lake Michigan. The cleanup effort has been estimated to have cost almost one billion dollars. In June and December 2010, two pipeline breaks resulted in 1,200 barrels of crude oil being leaked into Red Butte Gardens in Salt Lake City. In June of the following year 42,000 gallons/1,000 barrels of oil leaked into the Yellowstone River from a pipeline break associated with seasonal heavy flooding. Even more recently the same pipeline operator suffered a break on its 65 year old line in Arkansas releasing 500,000 gallons (11,900 barrels) of crude oil causing significant environmental damage in the process and posing a threat to the local water supply. While environmental damage can eventually be remediated or mitigated such is not the case when fatalities result from natural gas pipeline ignitions. In September 2010, eight people died when a natural gas pipeline located in the neighbourhood of San Bruno, California, ruptured and ignited. The root cause of this incident was attributed to a faulty long seam weld. This chapter presents an editorial view of essential elements for an effective D&D technology development program, and a compendium of challenges and needs from the perspective of technology end-users. The term “D&D” is used herein to describe a formal project management process that involves removal of real property from active or standby status and transition to a long-term end state (such as an industrial brown-field or an unrestricted green-field). Interim steps depend on the actual facility, and include characterization, deactivation, decommissioning, decontamination, demolition, waste management, and disposition or recycling of excess facilities and equipment. Life-cycle-cost analyses include surveillance and maintenance (S&M) while awaiting completion of D&D. This page is intentionally blank. The objective of this chapter is to provide some highlights on the requirements, analysis, and issues related to the seismic protection of piping systems. Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) Appendix A [1] contains General Design Criteria for Nuclear Power Plants. General Design Criterion 2, Design Bases for Protection Against Natural Phenomena under Group I of 10 CFR 50, mandates that structures, systems, and components (SSCs) important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches, without loss of capability to perform their safety functions. From the commissioning of the first commercial nuclear reactor more than 50 years ago, the nuclear power industry has been developing and improving reactor technology with particular emphasis on reliability and safety. There are several generations of reactors that have been developed or are being developed. These reactors are generally categorized as Generation I, II, III, III+, and IV reactors. Figure 15.1 illustrates the technology distilled from 50 years of successful nuclear operating experience that has led to the Generation III+ pressurized water reactors (PWRs). Nuclear energy plays a major role in meeting the world's energy needs. At the end of 2005, there were 443 nuclear power plants operating in 32 countries, with 25 more units under construction. These plants account for 17% of world's electricity. In a nuclear power plant, steam to turn the turbine and in turn the generator to produce electricity is produced through a controlled nuclear fission reaction. The steam producing part of a nuclear power plant, including the supporting systems, is called the nuclear steam supply system (NSSS). Beyond the NSSS, the remaining part of a nuclear power plant, the turbine/generator and remainder of the steam cycle, is called the balance of plant (BOP). In a boiling water reactor (BWR) NSSS, the steam is directly produced in the reactor pressure vessel (RPV). In the pressurized water reactors (PWRs), the steam is produced in steam generators that are connected to RPV. Both the BWRs and PWRs are classed as light water reactors (LWRs) since they use light water (as opposed to heavy water, D 2 O) as coolant and moderator. In the United States and many countries of the world, the RPV and the most of the NSSS components are designed, fabricated, tested, and inspected during service using the rules of ASME Boiler & Pressure Vessel Code. Specifically, Section II (for material selection and allowable stresses), Section III (for design and analysis), Section IX (for welding), and Section XI (for in-service inspection) of the Code are used. The overall objective of this chapter is to provide a description of the evolution of the BWR product line, including the current offerings (the Advanced Boiling Water Reactor, ABWR, and the Economic Simplified Boiling Water Reactor, ESBWR), along with the discussion on the role of ASME Code in the material selection, fabrication, design, and in-service inspection (ISI) of the BWR NSSS system. This chapter: (1) identifies the structural integrity issues in the ASME Boiler & Pressure Vessel Code, including Section II, Section III, Subsection NH (Class 1 Components in Elevated Temperature Service), Section VIII, and Code Cases that must be resolved to support licensing of Generation IV (Gen IV) nuclear reactors, if for very high temperature gas-cooled reactors; and (2) describes how the Code addresses these issues; and (3) identifies the needs for additional criteria to cover unresolved structural integrity concerns for very high temperature service. The U.S. Nuclear Regulatory Commission (NRC) license renewal process establishes the technical and administrative requirements for renewal of operating power plant licenses. Reactor operating licenses were originally issued for 40 years and are allowed to be renewed for up to an additional 20 years. The review process for license renewal applications (LRAs) provides continued assurance that the level of safety provided by an applicant's current licensing basis is maintained for the period of extended operation. The license renewal review focuses on passive, long-lived structures and components (SCs) of the plant that are subject to the effects of aging. The applicant must demonstrate that programs are in place to manage those aging effects. The review also verifies that analyses that are based on the current operating term have been evaluated and shown to be valid for the period of extended operation. As of March 2013, the NRC has renewed the licenses for 73 reactors, applications to renew the licenses of 15 additional reactors are under review. If the applications currently under review are approved, approximately 80% of the licensed operating reactors will have extended their life span by up to 20 years. As license renewal is voluntary, the decision to seek license renewal and the timing of the application is made by the licensee. However, the NRC expects that, over time, essentially most of the remaining U.S. operating reactors will request license renewal. The objective of this Chapter is to provide some details of many and sometimes unique ways in which the provisions of Section III and Section XI have been used in addressing the service-induced degradations in the BWR vessels, internals, and pressure boundary piping. Among the items covered are reactor internals, weld overlays, and reactor vessel. The most common form of service-induced cracking in the stainless steel and Ni-Cr-Fe components in the BWR pressure boundary is typically due to intergranular stress corrosion cracking (IGSCC). 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. Over the past 20 years, nuclear power plant design and operating practices have entered into an evolutionary period of change from the experience-based, deterministic requirements established by expert opinion to one that also includes a formalized application of risk importance determinations. This chapter traces the development of nuclear power plant probabilistic risk assessment (PRA) from its initial evolution as a means for evaluation of public safety through the recognition of its use to identify important safety concerns and ultimately to its use in helping to determine risk importance and the appropriate allocation of resources and in-service activities. It also discusses the current status of Codes and Standards that provide rules and guidance for the development of probabilistic risk analysis to support nuclear power plant applications. The Committee that has evolved into B16 [1] started life as a group of different organizations. All were becoming active in the early 1900's. This is the period when the ASME was asked and began to publish the Boiler and Pressure Vessel Codes. All of this activity was the result of the horrendous problems of boiler explosions that had wrecked the nation in the period after the Civil war. The technology for pressure equipment design continues to advance each and every day. The American Society of Mechanical Engineers (ASME) has been keeping pace with these advances over the last 100 years. Specifically, the ASME Boiler & Pressure Vessel Code, Section VIII, Division 1 [1] has become the International benchmark standard for pressure vessel design for a multitude of industries. In response to these ever-growing industries and their unique design requirements the document has grown in scope, technology, and size. Organization, maintenance, and use of a document of such magnitude have become a real challenge, especially with a writing style and format that is not readily adaptable to change. This chapter provides commentary into what are generally thought to be the more relevant issues of ASME Section VIII, Division 1 that need attention to better serve its users and to provide recommendations to resolve these issues. The 1960s and 1970s was a good time for the nuclear industry. The building of nuclear power plants was at an all-time high in the United States with the majority of the current operating fleet coming on-line in the late 1960s and early 1970s. While these plants were being built, the need to maintain these plants in a safe operating condition was discussed. The practice of performing inspections for fossil-fueled power plants was well established and the risk of operating fossil-fueled plant was understood. Although the nature and extent of the inspection share, in a large measure, evolved from several decades of operating experiences of many power stations, no formalized codes for conducting these inspections had been developed in the United States. Rather, a variety of rules were being imposed by the enforcement authorities having jurisdiction at the power plant sites. The ASME BPV Code promotes the introduction of new materials. As materials are produced to more closely fit users' requirements, new issues arise. One example is the introduction of “advanced” alloys into the B&PV Code, which represents a significant change in philosophy regarding materials' use at elevated temperatures. The inevitable acceptance of these alloys has prompted the introduction of new rules in the construction codes and new protocols in SC II that are more prescriptive than the rules and protocols adopted to govern the use of the more “traditional” materials. In view of the very significant changes in the power industry that have occurred over the last 30 years, this general trend is likely to continue. The following represent some of the lessons learned. This chapter summarizes past and ongoing developments in flaw Acceptance Standards and analytical evaluation procedures in Section XI of the ASME B&PV Code for nuclear piping degraded by flaws and corrosion processes that are in response to industry needs and lessons learned. A significant number of revisions have been made to address intergranular stress corrosion cracking in Boiling Water Reactor Class 1 austenitic piping, and primary water stress corrosion cracking in Alloys 82, 182, and 132 dissimilar metal welds in piping in Pressurized Water Reactors. More recent revisions to Section XI, as well as developments that are under review in Section XI, to address stress corrosion cracking are described in this chapter. A major revision to the flaw Acceptance Standards was to prohibit application to planar surface flaws that are detected in piping materials that are susceptible to stress corrosion cracking and are in contact with reactor coolant environments. This chapter focuses on the historical development of pressure relief, from its early days to present, as well as the application of pressure relieving requirements and pressure relieving devices (PRDs). Its purpose is not to re-cite ASME code rules, per se, as there are many books in circulation which cover that subject very well [1,2], but to provide designers and owners an insight on applying code rules more effectively for pressure equipment. While the focus is toward pressure vessels, other pressure equipment will be addressed such as boilers, piping, and tankage. Nuclear application is not discussed. This chapter provides a discussion of probabilistic risk assessment, its origin and evolution, and insights from its usage in the nuclear power industry. It provides a comprehensive set of references on various topics related to PRA basics and applications. These may be used by the reader for delving into the items discussed in greater detail. 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. Over the past 40 years piping designed to either the ASME B31.1 Power Piping Code or the ASME B31.3 Process Piping Code has proved it will perform well. That being said, why are there still failures in the industry? This chapter will attempt to review some significant failures or more common types of failures and provide some guidance on how they can be avoided. The design rules have proved adequate and have significant margin for safety if all of the design requirements have been identified. This is proving to be no small task and involves the original design as well as operations and maintenance activities. NuScale Power's unique, scalable approach to the deployment of commercial nuclear plants provides the financial, schedule, supply chain, and workforce advantages that will broaden the market for commercial nuclear power. NuScale Power also offers a new level of safety that may simplify siting and increase public acceptance of nuclear power. Although novel in its deployment approach, the factory-fabricated NuScale Power Module (NPM) relies on well-known light-water reactor technology, significantly reducing technological and regulatory risks. Most importantly, the NuScale design has achieved the “Triple Crown of Nuclear Safety.” That is, a NuScale reactor can safely shutdown and self-cool for an indefinite period of time with no operator action, no AC power or DC power, and no additional water. The design also eliminates the possibility of many accidents from occurring and adds greater defense in depth to protect against even the most severe events such as occurred at the Fukushima Daiich plant in Japan. These remarkable advances in safety are achieved while offering all the financial benefits that come with small modular infrastructure. The fundamental principles of hydrogen fusion make this process seem inevitable as a method of commercial power production. It is a clean power process, producing no harmful byproduct, using an inexhaustible supply of fuel, and hydrogen fusion is the source of all energy release in the larger universe, illumining countless galaxies with starlight. On a terrestrial, non-universal scale, hydrogen fusion can be accomplished on a table-top using a number of elementary procedures. This back matter contains the index.

This comprehensive reference contains contributions by over 50 experts from around the world. Topics cover aspects of power generation from all known sources of energy around the globe, including solar, wind, hydro, tidal and wave power, bio energy (including bio-mass and bio-fuels), waste-material, geothermal, fossil, petroleum, gas and nuclear. Nanotechnology and the role of NASA in photovoltaic and wind energy are also covered. A unique aspect of this publication is its foundation in scholarly discussions and expert opinions, enabling the reader to make decisions regarding which energy source(s) may be used in a given situation. The handbook, with nearly 700 pages, includes about 1,250 references and over 750 figures, tables and pictures as well as an extensive index. A brief biographical sketch is also included for each contributing author. This front matter contains the dedication, acknowledgements, contributor biographies, table of contents, preface and introduction. Click on the PDF icon to view the complete front matter. The use of solar energy for generating power is a concept that has been around for a long time. Some early examples of this, to be dealt with in the sections that follow, include dishes and troughs that were developed in a flurry of creative activity between the late 1800s and the early 1900s. There was somewhat of a hiatus from much development in first half of the 20th century until work again resumed after the Second World War. This was driven by interest in exploring space as well as other thrusts. An excellent summary of the early history of solar applications has been given by Butti and Perlin [1]. Momentum to examine renewable energy generally picked up considerably with the oil embargo of the 1970s. At that time, President Jimmy Carter outlined the challenge the US particularly, and the world more generally, faced in terms of dealing with energy issues. He warned that the situation was “the moral equivalent of war” in a 1977 speech. In this same year, he was responsible for forming the US Department of Energy, which combined some existing agencies, including the US Energy Research and Development Administration. He also initiated the Solar Energy Research Institute, in Golden, Colorado, that later became the National Renewable Energy Laboratory (NREL). Following that time, interest again waned as a result of falling energy prices. President Ronald Reagan removed the solar panels that were placed on the White House under President Carter. Oil again became the primary focus in energy. Various starts and stops to solar development followed, including the start of the large development of the Luz systems in the Mohave Desert area of California, which was later stopped when existing tax credits went away. Strictly speaking, all the practical energy sources for applications to human activities on the Planet Earth are from the sun. Even fossil fuels resulted from millions and millions years of layered deposition of once living plants (and animals eating plants), which obtained their energy intakes and conversion directly and indirectly from solar rays and reached their maturity before they were buried. From ancient times to today, humans have used various ways to harness direct solar radiation, from using it as a heating source to today's electricity generation systems. Solar energy is also responsible for such renewable energy sources as wind and wave power, hydroelectricity and biomass. Today, the total energy consumption in the United States still predominantly comes from fossil fuels, although recent interests in investing in wind and solar electricity have been accelerating. Solar thermal power plants, due to their capacity for large-scale generation of electricity and the possible integration of thermal storage devices and hybridization with backup fossil fuels, are meant to supply a significant part of the demand in the countries of the solar belt [1]. Nowadays, the high-temperature thermal conversion of concentrated solar energy is rapidly increasing with many commercial projects taken up in Spain, USA, and other countries such as India, China, Israel, Australia, Algeria, and Italy. This is the most promising technology to follow the pathway of wind energy in order to reach the goals for renewable energy implementation in 2020 and 2050. India has a population of 1.1 billion people (1/6th of the world population) and accounts for less than 5% of the global primary energy consumption. India's power sector had an installed capacity of 159,650 MW [1] as on 30th April, 2010. The share of installed capacity from different sources is shown in Fig. 4.1. The annual generation was 724 Billion units during 2008–2009 with an average electricity use of 704 kWh/person/year. Most states have peak and energy deficits. The average energy deficit is about 8.2% for energy and 12.6% for peak [1]. About 96,000 villages are unelectrified (16% of total villages in India) and a large proportion of the households do not have access to electricity. India's development strategy is to provide access to energy to all households. Official projections indicate the need to add another 100,000 MW within the next decade. The scarcity of fossil fuels and the global warming and climate change problem has resulted in an increased emphasis on renewable energy sources. India has a dedicated ministry focussing on renewables (Ministry of New and Renewable Energy, MNRE). The installed capacity of grid connected renewables is more than 15,000 MW. The main sources of renewable energy in the present supply mix are wind, small hydro- and biomass-based power and cogeneration. In 2010, India has launched the Jawaharlal Nehru Solar Mission (JNSM) as a part of its climate change mission with an aim to develop cost-effective solar power solutions. Solar energy can trace its roots to the early 19th century, when in 1838 French physicist Edmund Becquerel [1],[2] published his findings about the nature of materials being able to turn light into energy. He discovered the photovoltaic effect while experimenting with an electrolytic cell made up of two metal electrodes. Becquerel found that certain materials would produce small amounts of electric current when exposed to light. At the time this was an interesting discovery that was not appreciated. Twenty years passed before Auguste Mouchout [1], a French mathematics teacher, designed and patented the first machine that generated electricity using the sun. Mouchout began his work with solar energy in 1860. He produced steam by heating water using a glass-enclosed, water-filled iron cauldron. Mouchout then added a reflector to concentrate additional radiation onto the cauldron, thus increasing the steam output. He succeeded in using his apparatus to operate a small steam engine. At the 1878 Paris Exhibition, he demonstrated a solar generator that powered a steam engine, similar to the one shown in Figure 5.1. This engine included a mirror and a boiler that drove an ice-maker that produced a block of ice. Later in 1869, Mouchot wrote one of the first books devoted to solar energy: “Le Chaleur Solaire et les Applications Industrielles.” Mouchout's work help lay the foundation for our current understanding of the conversion of solar radiation into mechanical power driven by steam. Since the beginning of NASA over 50 years ago there has been a strong link between the energy and environmental skills developed by NASA for the space environment and the needs of the terrestrial energy program. The technologies that served dual uses included solar, nuclear, biofuels and biomass, wind, geothermal, large-scale energy storage and distribution, efficiency and heat utilization, carbon mitigation and utilization, aviation and ground transportation systems, hydrogen utilization and infrastructure and advance energy technologies such as high-altitude wind, wave and hydro, space solar power (from space to earth) and nanotech photovoltaics. NASA, in particular with wind and solar energy, had extensive experience dating back to the 1970's and 19 80's and continues today to have skills appropriate for solving our nation's energy and environmental issues that mimic in fact those needed for space flight. NASA has long been interested in maturing new laboratory level technologies into industry products and has a well-founded system of interactions with Universities for research leading to Proof-of-Concept and then to prototype demonstrations and mission applications with industry. The capability to assess and then test the feasibility for future commercial development is a well-honed NASA skill. This has certainly promoted new businesses and job growth as a result. That is particularly evident in the case of photovoltaic energy but also relevant for wind energy as well. The energy from the wind has been harnessed since early recorded history all across the world. There are proofs that wind energy propelled boats along the Nile River around 5000 B.C. The use of wind to provide mechanical power came somewhat later in time — by 200 B.C. simple windmills started pumping water in China, and vertical-axis windmills with woven reed sails were grinding grain in the Middle East. The Europeans got the idea of using wind power from the Persians who introduced it to the Roman Empire by 250 A.D. By the 11th century, a strong focus on technical improvements enabled wind power to be leveraged by the people in the Middle East extensively for food production. Returning merchants and crusaders carried this idea back to Europe where the Dutch refined the windmill and adapted it for draining lakes and marshes in 1300s. In the late 19th century settlers in America began using windmills to pump water for farms and ranches, and later, to generate electricity for homes and industry applications. Although the industrial revolution influenced the propagation of wind energy, larger wind turbines generating electricity continued to appear. The first one was built in Scotland in 1887 by prof. James Blyth from Glasgow. Blyth's 33-ft. tall, cloth-sailed wind turbine was installed in the garden of his holiday home and was used to charge accumulators that powered the lights, thus making it the first house in the world to have its wind power supplied electricity. At the same time across the Atlantic, in Cleveland, Ohio, a larger and heavily engineered machine was constructed in 1888 by Charles F. Brush. His wind turbine had a rotor 17 m in diameter and was mounted on an 18-m tower. Although relatively large, the machine was only rated at 12 kW. The connected dynamo had the ability to charge a bank of batteries or to operate up to 100 incandescent light bulbs, three arc lamps, and various motors in Brush's laboratory. The machine was decomissioned soon after the turn of the century. In the 1940s, the largest wind turbine of the time began operating on a Vermont hilltop known as Grandpa's Knob. This turbine, rated at 1.25 MW fed electric power to the local utility network for several months during World War II. In Denmark, wind power has played an important role since the first quarter of the 20th century, partly because of Poul la Cour who constructed wind turbines. In 1956, a 24-m diameter wind turbine had been installed at Gedser, where it ran until 1967. This was a three-bladed, horizontal-axis, upwind, stall-regulated turbine similar to those used through the 1980s and into the 1990s for commercial wind energy development, see Fig. 7.1. The popularity of using the wind energy has always fluctuated with the price of fossil fuels. When fuel prices fell in the late 1940s, interest in wind turbines decreased, but when the price of oil skyrocketed in the 1970s, so did worldwide interest in wind turbine generators. Humans have been harnessing the energy in the wind for several thousand years. Early uses included sailboats and windmills used to pump water and grind grain into flour — hence the term wind mill . In the 1800s, settlers in the western United States used wooden windmills to pump water, and many are still standing. In the late 1800s, the windmill was connected to an electric generator to produce electricity — hence, wind turbine . In the 1970s and 1980s, wind turbines were clustered into wind farms and connected to the electric grid in California, which marked the first commercial, utility-scale use of wind energy. The size of those wind turbines were 100 kW and smaller. In the following decades, wind turbine technology progressed quickly, and by 2010, grid-connected wind turbines were typically in the 2-MW range and turbines as large as 6-MW have been deployed [17]. Adequate wind speeds are essential to the success of wind energy facilities. The potential energy in wind is determined as a function of the cube of the speed. Given the cubic relationship between power and wind speed, when wind speed doubles there is eight times more power available. Interest in the application of modern wind energy grew in the Netherlands in the 1970s when the limit of fossil fuels became clear. Wind energy has been an important source of energy in The Netherlands for centuries and the country was known for the many wooden wind mills especially in the coastal regions. At the same time, research activities in the modern wind energy were started, which have led to the relatively large wind energy research community in the Netherlands today. Wind energy research activities in the Netherlands are predominantly performed at the Energy research Centre of the Netherlands (ECN) and the Delft University of Technology (DUT). Both institutes are involved in wind energy research since the start of the modern wind turbines, the 1970s. These institutes match their research programs with each other as close as possible. ECN Wind Energy has a research staff of 55 scientists and DUT has a research staff of 15 permanent researchers and more than 35 PhD students. Another institute dedicated to wind energy research is the foundation Knowledge Centre WMC that has been founded by the DUT and ECN in 2003 with an additional research staff of 25 scientists. Although the major part of the wind energy research is concentrated in these institutes, many other universities and scientific institutes contribute to the research with dedicated and specialized research. The current wind energy research and associated industrial activities are taking place in an international context, mostly the European context, therefore the research activities not only take account of the long-term energy research program of the Dutch government, such as the long-term energy research program EOS [1], but also of the R&D priorities defined in the international context, such as the Strategic Research Agenda (SRA) of the wind energy sector [2]. The three wind energy research organizations are well represented in international bodies such as European Wind Energy Association (EWEA), European Academy of Wind Energy (EAWE), International Energy Agency (IEA), International Electrotechnical Commission (IEC), International Network for Harmonised and Recognised Measurements in Wind Energy (MEASNET), European Wind Energy Technology Platform (TPWind) and the European Energy Research Alliance (EERA). Renewable energy technologies except for large hydroelectric power have, until recently, played a marginal role in most of the countries in terms of energy supply mix and it is no different in developing nations such as India. Global concerns about excessive consumption and mitigation efforts have put Renewable Energy Technologies on centre stage not merely because of their novelty but out of necessity. In this paper we attempt to put in perspective the opportunities and challenges that these emerging technologies face in a wide range of situations afforded by different countries in the region. Approach taken by different countries is analyzed with a peep through the recent historic information and its influence on present and future developments. An attempt is made to estimate future role of the renewable energy technologies in the energy market. The development of dams on rivers, with associated benefits of water storage for flood control, irrigation, and “hydropower” has played a vital role in advancing civilization throughout history. Of these, hydropower ingeniously and yet so simply combines two of the most fundamental components of nature on planet Earth — water and gravity — to help sustain our survival and improve our lifestyle. This chapter will describe the role of hydropower from past to present and potentially into the future. Hydropower will be demonstrated to be a safe, reliable, and renewable energy resource worldwide, essential to our overall power and energy mix, both traditionally from rivers, through recent and growing development of pumped energy storage from lower to upper water reservoirs and evolving in the future with tidal and wave energy from the oceans. Power generation in India has come a long way from about 1000 MW at the time of independence (August, 1947) to about 160,000 MW as on 31st March 2010 (end of Financial Year). The share of hydro power in this growth, in these over six decades, has also been impressive as it increased from about 500 MW at the time of independence to about 37,000 MW as of March 2010. But the present level of hydro power exploitation is only about 25% of the ultimate installed capacity estimated at 150,000 MW. The demand of power is increasing rapidly (at the rate of over 8% per annum) so has been the realization that hydro power energy, being a renewable source, be exploited to its full available potential. In this endeavor, of the balance over 113,000 MW of power, i.e., yet to be commissioned, projects with about 14,000 MW are under construction and about 100,000 MW are under various stages of implementation. The hydro project implementation and ownership remained with the State Governments or with the Power Corporations owned by States or Central Government for about four and a half decades. Due to slow pace of development and also in line with international stress for liberalization of economy, the Government of India reviewed the policy of power development. In 1992, the power sector was opened up allowing private capital participation in its development along and in parallel with continued development under public sector (under the Five-Year Plan system) [1]. With inherent constraints associated with hydro power development, construction activity in the private sectors did not show expected results in the initial period of over one decade. But it is picking up now and at present projects aggregating to about 25,000 MW are allocated to private developers on build, own, operate, and transfer basis (BOOT). Of these, projects with installed capacity of 20,000 MW are under various stages leading to their implementation including survey investigation, DPR preparation and clearances, about 4000 MW are under construction and over 1400 MW are in operation. Power generation from waves and tidal currents is a nascent industry with the potential to make globally significant contributions to renewable energy portfolios. Further development and deployment of the related, immature technologies present opportunities to benignly tap large quantities of renewable energy; however, such development and deployment also present numerous engineering, economic, ecological, and sociological challenges. A complex research, development, demonstration, and deployment environment must be skillfully navigated if wave and tidal power are to make significant contributions to national energy portfolios during the next several decades. Oil shortages in the 1970s stimulated the development of new technologies to convert biomass to heat, electricity, and liquid fuels. Combined firing of biomass with coal reduces emissions, provides opportunities for high efficiency, and reduces fossil carbon use. Cofiring and markets for transportation fuels have stimulated global trading in biomass fuels. Pyrolysis of biomass to liquid fuels creates opportunities for co-products, such as biochar, which can help sequester carbon and offset emissions from fossil fuels. This chapter provides an overview of biomass fuels and resources, the biomass power industry, conventional and new technologies, and future trends. Advances in combustion and gasification are described. Thermal and biological conversions to liquid fuels for heat, power, and transportation are described with implications for stationary use. The generation of waste, whether industrial or residential, is a fact of life in our society today. Nearly everything we do creates some type of waste. It is estimated that the United States alone generates 7.6 billion tons per year of non-hazardous industrial waste [1], 48 million tons per year of hazardous waste [2], and 250 million tons per year of municipal solid waste (MSW) [3]. Many of the waste streams and industrial by-products we generate each year contain recoverable energy. Capturing and utilizing this energy can create a positive impact both economically and environmentally. It not only extends a material's life cycle, but it also reduces the volume of waste sent to landfills, conserves non-renewable resources, and helps reduce manufacturing costs by providing a lower cost alternative to the rising costs of energy and waste disposal. Depending on the waste and the fossil fuel it is replacing, it may even help reduce our carbon footprint. The volume of waste we generate continues to grow each year. In the United States, MSW generation has increased from 88 million tons in 1960 to 250 million tons in 2008 [4]. The amount of industrial waste has grown as well. Some of this increase is due to the fact that we have 120 million more people in the United States today than we did 50 years ago, but much of it is a result of our changing lifestyles and consumption habits. Today, we use significantly more disposable items than we did 50 years ago, and we have developed thousands of new chemicals, plastics, paints, and adhesives; all of which generate their own production by-products that need to be disposed of. Figure 15.1 shows the growth in MSW generation rates from 1960 to 2008. Geothermal energy constitutes an indigenous, sustainable, continuous, base load renewable resource available to power developers on most continents. This chapter discusses geothermal energy basics, resource exploration, and types of resources along with their utilization, sustainability, benefits, and the potential environmental consequences of resource development. The current state of knowledge and possible expansion of the resource base via Enhanced Geothermal Systems technologies are also discussed. Advanced Ultra Supercritical (A-USC) is a term being used to describe a coal fired power plant design with the inlet steam temperature to the turbine at 700°C to 760°C (1292°F to 1400°F). Nickel alloy materials are required. The term Ultra Supercritical (USC) is a term for plants currently designed and operating at 600°C (1112°F) using available and suitable ferritic and stainless steels. Increasing efficiency of the Rankine regenerative-reheat steam cycle to improve the economics of electric power generation and achieve lower cost of electricity has been a long sought after goal. Efficiency has more recently been recognized as a means for reducing the emission of carbon dioxide and its capture costs, as well as a means to reduce fuel consumption costs. Programs have been established by nations, industry support associations and private companies to advance the technology in steam generator design and materials development of nickel based alloys needed for use above 700°C. The worldwide abundance of less expensive coal fuel has driven economic growth. The challenge is to continue to advance the improvement of efficiency for coal fired power generation technology, representing nearly 50% of the United States' (U.S.) production, while helping to maintain economic electric power costs with plants that have favorable electric grid system operational characteristics for turndown and rate of load change response. The Newcomen steam engine operated at about 0.5% thermal efficiency in 1750 [1]. Major efficiency milestones such as James Watt's 1769 patented improvement to a Newcomen steam engine by adding a separate condenser is credited with achieving 2.7% thermal efficiency by 1775 and became a major propellant of the Industrial Revolution due to the economic benefit of the fuel savings attained. Watt's 1782 patent for expansive working, double acting cylinder engine, is credited with achieving 4.5% efficiency by 1792. James Watt, however, is also assailed for causing delay in economic advancement due to his reluctance to raise the working pressure of the steam engine. Lacking in good boiler instrumentation such as reliable water level and pressure gages, Watt's concerns were probably well founded in limiting the operating pressure to about 0.034 MPa (5 psig). After Watt's patents began to expire, Richard Trevithick is credited with engine improvements that permitted increasing steam pressure, to 1MPa (145 psig), to achieve 17% thermal efficiency by 1834. Changing climate and rising carbon dioxide (CO2) concentration in the atmosphere have driven global concern about the role of CO2 in the greenhouse effect and its contribution to global warming. Since it has become widely accepted as the primary anthropogenic contributor, most countries are seeking ways to reduce CO2 emissions in an attempt to limit its effect. This effort has shifted interest from fossil fuels, which have energized the economies of the world for over a century, to non—carbon-emitting or renewable technologies. For electricity generation, the non-carbon technologies include wind, solar, hydropower, and nuclear, whereas low carbon technologies use various forms of biomass. Unfortunately, all current options are significantly more expensive than current commercial fossil-fueled technologies. Although use of wind and solar for power generation is increasing, they are incapable of supplying base load needs without energy storage capacity which is currently impractical at the scale required. The only technologies capable of sustaining base load capacity and potential growth are coal, natural gas, and nuclear. Nuclear has a long lead time to commercial operation, and concerns about long-term disposal of waste have not been resolved. Natural gas, although lower in carbon emissions than coal, still produces CO2, and its availability and price have proven to be volatile. Although recent discoveries have significantly increased availability of natural gas reserves, the pricing impact of higher cost extraction methods is not yet known. In many countries, including the United States, coal is an abundant and low-cost source of fuel for generating electricity. Figure 18.1 shows the International Energy Agency's (IEA's) global electricity-generating capacity by fuel in 2007 and that projected for 2030. It should be noted that the current IEA forecast indicates more than a 70% increase in coal-based power generation by 2030. In the United States, coal currently fuels about 50% of power generation and represents an enormous infrastructure investment. Consequently, it will take considerable investment to move away from this base load fuel. The economic behavior of all fuels share many common characteristics and it is important for the reader to apply principles concerning one fuel to others as the need arises. In many cases, we use the United States as an example. This is because there is better long-term data available than for other countries. In addition, it is a reference point to compare countries especially when evaluating the impact of policies. When comparing fuels we will focus on the cost per Btu and Btu/lb and Btu/ft3. Efficiency in converting Btu's to mechanical or electrical forms of energy is important as well. The scale of the “conversion plant” may affect efficiency and fuel use for a particular purpose. Fossil fuels supply almost all of the world's energy and feedstock demand. Among the fossil fuels, coal is the oldest, most abundant, and widely available form of fossil fuel. Coal constitutes over 75% of the world's fossil fuel. Since the beginning of the industrial revolution, coal has been the backbone of the world energy system. Before the discovery of oil and gas, coal was used to generate town gas via gasification for lighting and heating purposes. During World War II, coal gasification technology was extensively used in Germany to produce oil substitutes. After World War II, oil replaced coal as the major source of energy, and the interest in coal gasification remained dormant until the recent rise in energy prices. Since 2000, several gasification plants have become operational, most of these plants are for the production of chemicals and only a handful of plants for power generation. Coal is among the cheapest fuel available. Figure 20.1 compares the historical and projected price of major energy sources in the United States [1]. Projections of the Energy Information Administration (EIA) indicate that coal is likely to remain the cheapest fuel in the foreseeable future and will likely become cheaper as the price of other preferred energy sources rise. Based on the energy content in terms of price per Btu, electricity commands the highest premium, followed by oil and gas. This price differential is a key driver that determines the interest in coal and its conversion into other energy substitutes. During the 1960s and 1970s, when most nuclear plants in the United States were being designed and constructed, costs escalated in an uncontrolled manner. This was due to many factors, and all interested parties had an influence on the results. This chapter will address these issues, not to accuse or place blame, but to analyze and discuss in order to learn from the past. The first commercial nuclear power plants were designed and constructed in the mid 1950s and early 1960s. Because these plants were based on the new technology of using nuclear reactions to heat water, thus producing steam to turn the turbines, there were no Codes or Standards that specifically addressed nuclear power plant components. Therefore, these early plants used the most applicable Codes and Standards that were available at the time [1–3]. That was the proper course of action. At the time, the question was, what was available in terms of Codes and Standards? What Codes and Standards were really appropriate for pressure vessels and piping in a nuclear power plant? The designers of some nuclear power plant systems considered that question and determined that the nuclear reactor that contained the nuclear fuel was equivalent to a boiler in a fossil fuel power plant. Other designers felt that the nuclear reactor was an unfired pressure vessel. The available Code for the design and construction of the boiler were the American Society of Mechanical Engineer's Boiler and Pressure Vessel Code, Section I, “Power Boilers” and Section VIII, “Unfired Pressure Vessels.” [1, 2]. Both approaches were correct because the differences between the two ASME Codes were minimal for the design and construction of the pressure vessel. The appropriate Piping Code used was ASME B31.1, “Power Piping” [3]. Both Section I and Section VIII were used for the first nuclear power plant reactor pressure vessels, and B31.1 was used for the associated pressure piping, regardless of whether Section I or Section VIII was used for the nuclear reactor pressure vessel. In the 1960s, nuclear power for electricity production was one of the greatest future economic promises, both in the United States and internationally. Electricity “too cheap to meter” was the claim, and US utilities and supplier corporations jumped in wholeheartedly. Yet 50 years later, the industry is just pulling out of decades-long doldrums. No new nuclear plants have gone on line in the United States for over 30 years. All US Nuclear Steam System Suppliers (NSSSs) are either foreign owned or have major overseas partners, and we no longer have the domestic infrastructure and capacity to produce the components needed for new plants. The current nuclear fleet, although operating safely and economically, generates only about 20% of US electricity. Given shortages of domestic fossil fuels and concerns about greenhouse gases and global warming, not taking greater advantage of this United States—developed energy resource defies logic. What happened? And what can the industry do to keep this from reoccurring during the current reemergence of nuclear power as a major clean energy source of the future? Many of the problems were no doubt political and beyond the scope of this technical article. Yet most were self-inflicted, many of them related to component mechanical and structural integrity that can and should be avoided in the future, if we can learn from the lessons of the past. The purpose of this article is to highlight the issues that the authors have personally seen and been involved in, as engineers and technical managers in the nuclear industry for over 40 years. The authors' backgrounds combine employment at major nuclear vendors during the formative days of the industry, plus utility engineering perspective provided by a co-author who was a utility participant and ultimately chairman of the industry's generic Materials Reliability Program. The article looks retrospectively at the root causes of these problems, itemizes the lessons learned, and recommends an approach going forward that will anticipate and hopefully prevent the industry from repeating these mistakes in the future. In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. Six reactor concepts were chosen for further development: the sodium fast reactor (SFR), the very-high-temperature gas-cooled reactor (VHTR), the lead or lead-bismuth cooled liquid metal reactor, the helium gas-cooled fast reactor, the molten salt reactor (MSR), and the super critical water reactor. In view of sustainability, the Generation IV reactors should not only have superior fuel cycles to minimize nuclear waste, but they should also be able to produce process heat or steam for hydrogen production, synthetic fuels, refinery processes, and other commercial uses. These reactor types were described in the 2002 Generation IV roadmap. Different projects around the world have been started since that time. The most advanced efforts are with reactors where production experience already existed. These reactors include the SFR and the VHTR. The other reactor types are still more in a design concept phase. This chapter briefly describes the six Generation IV concepts and then provides additional details, focusing on the two near-term viable Generation IV concepts. The current status of the applicable international projects is then summarized. These new technologies have also created remarkable demands on materials compared with light water reactors (LWRs). Higher temperatures, higher neutron doses, environments very different from water, and design lives of 60 years present a real engineering challenge. These new demands have led to many exciting research activities and to new Codes and Standards developments, which are summarized in the final sections of this chapter. In considering the future of nuclear power for electricity production in the U.S., it is necessary to consider the present public perception of nuclear power. It is also necessary to consider public perceptions of the various competing sources of electricity production. These include coal, natural gas, and the several “green” or “renewable” sources, including hydro, wind, and solar. Note that petroleum is no longer part of this discussion; its use has diminished to barely 1% of electricity production. Chapter 24 makes extensive use of referenced publications of the Nuclear Energy Institute, to take advantage of their expertise on specific subjects. This chapter also makes extensive use of referenced quotations from the works of established authors in the field to minimize misinterpretation of their work. In 1994, about the time that the last nuclear power plant was completed, USA Today announced [1]: “Essentially, the nation decided that nuclear power wasn't worth the price.” …“Nuclear energy provides only 21% of our electrical needs, 4% of our total energy consumption. Is it worth the trouble? Not until the problems of waste, safety and cost are completely resolved.” The fundamental design concepts embodied in steam turbines and generators are fairly simple—allow high pressure steam to expand through a series of blades mounted to a rotating member to extract thermal energy from pressurized steam and convert it to mechanical energy in the form of torque. Then utilize this torque to rotate a large electromagnet past a series of conductors to produce current flow in the conductors—pretty simple. However, as simple as this seems, these are very complicated machines that involve essentially all aspects of engineering—statics and dynamics, heat transfer, fluid mechanics, thermodynamics, ferrous and nonferrous metallurgy, organic and inorganic chemistry, steam cycle chemistry, materials behavior (stress/strain), fracture mechanics, corrosion, erosion, electrical (power) engineering, electric circuits and circuit models, electromagnetics, control theory and controls, power system analysis, dielectrics and electrical insulation, tribology, and various forms of materials joining from glues and resins to soldering, brazing, and welding. Steam turbines have historically been the prime source of power for electric power generation. Turbines come in a variety of types with regard to inlet and exhaust steam conditions, casing and shaft arrangements and flow directions. This chapter will focus on steam turbines currently being applied to power generation. The steam conditions will include those currently applied to fossil fired power plants, combined cycle and nuclear power units. Currently, fossil fired plants, whether coal fired or gas fired, typically supply steam at 1800 to 3500 psig steam pressures with 950 to 1050°F main steam and reheat temperatures. Double reheat units are few and will not be discussed in this chapter in much detail. Combined cycle plants (CC) typically have multiple drum heat recovery steam generator (HRSG) with inlet temperatures of 1050°F for main steam and reheat steam (see Chapter 27). While no new nuclear units have been installed in recent years, plans are underway to apply the AP1000 nuclear systems (see Chapters 23 and 24). Past steam conditions were typically saturated steam (0.2% moisture) and reheat temperature of 500°F. All current applications are condensing designs with regenerative extraction, except for CC units. Current application can be used up to 8 to 10 inHgA exhaust pressure. The application of air cooled condensers is more prevalent in CC applications. Current designs are axial flow turbines. However, in some applications, radial flow stages are used in the inlet stages. Typical arrangements are tandem compound. This means more than one turbine casing's rotors are coupled together on the same shaft. For example a turbine train consisting of a high-pressure turbine (HP), an intermediate-pressure turbine (IP) and two low-pressure turbines (LP) would have all of these in line and coupled to an electric generator. Cross compound units (two or more shafts) are currently in use. Some arrangements have the HP and IP on one shaft along with a generator and the LP turbines on a separate shaft along with another generator. In many instances the low-pressure turbine shaft is running at half speed and has larger annulus areas to reduce LP turbine exhaust velocity and associated leaving losses. However, due to increased blade height and associated exit annulus area for full speed units and the higher costs, cross compound units are not currently in favor. Casing configurations for fossil units are typically high-pressure unit, intermediate-pressure unit, and one or more low-pressure units exhausting to the condenser. The high-pressure and intermediate-pressure units are frequently combined into one casing (HP-IP). Thermal Power Plants are power plants operating on carbon-based fuel such as coal, natural gas, and petroleum products. There are five major types of thermal power plants in use on the power grid today: Steam Turbine Power Plants produce electric power by creating steam at high pressure and temperature (superheated steam at 2000 psia/138 bars and 1500°F/815°C) in boilers, which is then expanded through a steam turbine causing the turbine to drive a generator, which then produces electric power. The steam leaving the turbine is usually sent to a condenser, which maintains a vacuum and where the steam is condensed back to a liquid condensate (water). The steam turbine follows the Rankine Cycle. Steam turbine power plants can produce up to 2000 MW of power and are most widely used plant types in the world. These plants have a thermal efficiency between 28% and 35%. Simple Cycle Gas Turbine Power Plants are plants that follow the Brayton Cycle; in gas turbines, the air is compressed in the compressor section of the turbine to a high pressure (580 psia/40 bars) and temperature (1300°F/704°C), and in the combustor the air is further heated to a higher temperature (2600°F/1426°C) at constant pressure. The gas leaving the combustor is at a high pressure and high temperature and is then expanded through the turbine section. The gases leaving the turbine are at a pressure of about (15.0 psia/1.03 bars) and a temperature of about (1200°F/649°C). The turbine drives the gas turbine's compressor and the generator, which produces electric power. These gas turbines can produce up to 300 MW of power. The gas turbine power plant can be installed in 12–18 months, and thus have been used widely in developing nations where energy requirements change rapidly. These plants have a thermal efficiency between 25% and 45% depending on the size of the plant. Hydro Tasmania has developed a remote island power system in the Bass Strait, Australia, that achieves a high level of renewable energy penetration through the integration of wind and solar generation with new and innovative storage and enabling technologies. The ongoing development of the power system is focused on reducing or replacing the use of diesel fuel while maintaining power quality and system security in a low inertia system. In recent years Hydro Tasmania has undertaken several renewable energy developments on King Island with the aim to reduce dependence on diesel, reduce operating cost and greenhouse gas emissions, and demonstrate the potential for renewable energy penetration in power systems. This has been achieved through the substitution of diesel based generation with renewables such as wind and solar and the integration of enabling technologies such as storage and a dynamic frequency control resistor. The projects completed to date include: • Wind farm developments completed in 1997 and expanded to 2.25 MW in 2003; • Installation of a 200 kW, 800 kWh Vanadium Redox Battery (2003); • Installation of a two-axis tracking 100 kW solar photovoltaic array (2008), and • Development of a 1.5 MW dynamic frequency control resistor bank, that operates during excessive wind generation (2010). This chapter describes shell-and-tube and plate and frame types of power plant heat exchangers and tubular closed feedwater heaters and the language that applies to them. The chapter briefly discusses Header Type Feedwater Heaters and their application and use. It defines the design point used to establish exchanger surface and suggests suitable exchanger configurations for various design-point conditions and the criteria used to measure performance. It does not cover power plant main and auxiliary steam surface condensers because of the differences in how they are designed and operated The chapter briefly discusses the effects on the exchanger of normal and abnormal deviations from design point during operation. The subject of this chapter is the capital equipment in a steam power plant that is used to condense the exhaust steam from the lowest pressure turbine, by using water as the cooling medium. In the applied heat transfer literature, such heat transfer equipment is often simply referred to as the “Surface Condenser.” A surface condenser is necessarily a large piece of equipment because more than 60% of the thermal energy produced by a power plant ends up as low enthalpy (waste) heat. This is because of the inherent thermodynamic limitation of the Rankine Cycle, which must be rejected by the condenser to the environment. The heat transfer area in a power plant's surface condenser easily dwarfs that in any other heat exchanger in the plant. Classical thermodynamics holds that the lower the temperature of the heat sink, the higher the efficiency of the Carnot cycle. Therefore, attaining the lowest possible condensing temperature in the heat sink of the Rankine Cycle—the surface condenser—is a primary goal in surface condenser design. Since the saturation temperature and pressure of steam vary in a proportional manner at low pressures, the objective of low condensing temperature translates to that of a low condenser operating pressure. Accordingly, surface condensers are operated at as high a level of vacuum (typically 1.0 to 2.5 in. of mercury absolute) as the quantity and temperature of the cooling water would allow. The subatmospheric condition in the condenser and portions of turbine assembly and auxiliaries promotes leakage of air into the system. In boiling water type of nuclear plants the motive steam acquires additional non-condensibles due to radiological disassociation of water into hydrogen and oxygen. Whatever their origin, these so-called non-condensibles tend to collect in the lowest pressure region in the power cycle which is the steam space in the surface condenser. Unless removed continuously and efficiently, they may sharply interfere with the heat transfer process in the condenser. Kern [8] quotes observed data of Othmer which indicates that even 1% volumetric concentration of air in steam reduces the condensing film coefficient by approximately 45%. The oxygen in the non-condensibles is another source of concern. Oxygen is known to actuate corrosion of condenser internals. Austenitic stainless steel tubing can become highly susceptible to stress corrosion in the presence of even small concentrations of oxygen. Therefore, efficient collection and removal of non-condensibles is of paramount importance in surface condenser design. A reliable design for collection and expulsion of the non-condensibles is particularly important in the surface condensers used in geothermal plants where the steam extracted from the ground has a large fraction of associated gases. Industrial enterprises have significant negative impacts on the global environment. Collectively, from energy consumption to greenhouse gases to solid waste, they are the single largest contributor to a growing number of planet-threatening environmental problems. According to the Department of Energy's Energy Information Administration, the industrial sector consumes 30% of the total energy and the transportation sector consumes 29% of the energy. Considering that a large portion of the transportation energy costs is involved in moving manufactured goods, the energy consumption of the industrial sector could reach nearly 45% of the total energy costs. Hence, it is very important to improve the energy efficiency of our manufacturing enterprises. In this chapter, we outline several different strategies for improving the energy efficiency in manufacturing enterprises. Energy efficiency can be accomplished through energy savings, improved productivity, new energy generation, and the use of enabling technologies. These include reducing energy consumption at the process level, reducing energy consumption at the facilities level, and improving the efficiency of the energy generation and conversion process. The primary focus of this chapter is on process level energy efficiency. We will provide case studies to illustrate process level energy efficiency and the other two strategies. Chapter 32 discusses a variety of nano-coatings and materials used in the energy and power generation fields. Nano-coatings, nano-composite coatings, nano-layered coatings, functional graded coatings, and multifunctional coatings will be presented. These coatings can be deposited by a wide range of methods and techniques including physical vapor deposition processes (PVD) such as cathodic arc, sputtering, and electron beam evaporation as well as chemical vapor deposition (CVD) and thermal spray. The various types of nano-coatings and their roles in assisting to generate energy and power for the fuel cell, solar cell, wind turbine, coal, and nuclear industries will also be discussed. This chapter provides a brief description of how the past and present state-of-the-art nano-technology within the different industrial areas such as the turbine, nuclear, fuel cell, solar cell, and coal industries is used to improve efficiency and performance. Challenges facing these industries as pertaining to nano-technology and how nano-technology will aid in the improved performance within these industries will also be discussed. The role of coating constitution and microstructure including grain size, morphology, density, and design architecture will also be presented with regards to the science and relationship with processing-structure-performance relationships. This chapter will conclude with a summary of the future role of nano-technology and nano-coatings and materials in the fields of power generation and energy. This back matter contains the index

Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management, May 14–18, 2006, New Orleans, Louisiana. The objective of this conference is to promote rational decision making to assure safety and reliability and to optimize the use of resources for complex systems. This is to be achieved through the use of risk assessment and management methods. The conference provides a public forum for information exchanges covering both theory and applications of these methods to the design and operation of technological systems and processes. This front matter contains the table of contents In probabilistic safety analysis (PSA) of highly redundant systems common cause failures (CCF) of components in general contribute to a large extent to the systems unavailability. As CCF events are rare events, operating experience of “classes” of components of the same type (like centrifugal pumps or motor operated gate valves) combined from “similar” plants and systems is used to generate CCF data with sufficient statistical quality. In the German approach, the operating experience of all German nuclear power plants since the beginning of their operation is combined although these plants belong to different types and generations and although maintenance and operational practices have advanced. In order to take into account plant and system specific differences in technical and operational details, specific boundary conditions were introduced to be able to adjust the generic CCF data to component specific conditions in the analysed plant and system. Most of these boundary conditions consider modern maintenance strategies and surveillance methods which were introduced in the plants to allow an early detection of potential common cause failure mechanisms before these mechanisms develop to multiple failures. Other boundary conditions take into account system specific technical aspects like quality of the process medium which may have influence on the probability of occurrence of some types of CCF mechanisms. In a common effort of all German utilities and GRS the existing German CCF event data base recently was re-evaluated in order to introduce these aspects in the German CCF data pool. GRS discussed each of the originally approximately 250 CCF events in the pool with PSA, system and maintenance experts and operational staff from each German utility. Plant specific maintenance policies and specific system characteristics could be identified which would reduce or increase the probability of occurrence or the impact of a CCF mechanism. For each discriminating aspect a respective boundary condition was formulated. Then all involved experts firstly assessed the event as it occurred and secondly assessed a hypothetical event showing the same CCF mechanism under the assumption that the new boundary condition is effective. Thus, two quantitative event assessments based on the component impairment vector approach of the international CCF data exchange project ICDE were introduced in the CCF event data pool for one observed CCF event. The result of these discussions was that for about half of the events in the original CCF event data pool additional boundary conditions were defined. For some of these events up to four different boundary conditions were specified. In total there are now 321 event assessments in the German CCF event data pool. First experience with the new data pool could be gained by calculating CCF probabilities for different types of components with the coupling model. Compared to coarser former results, the results achieved now are more realistic as they differentiate more for specific characteristics of the analysed components. The paper describes the concept of boundary conditions and presents the common activity of the German nuclear power plant utilities and the GRS to set up the new CCF event data pool. Furthermore, some comparative results from the calculation of CCF probabilities for different boundary conditions are presented and an outlook on ongoing activities of the German utilities and GRS to update the CCF event data pool is given. A research project on the impact of CCF treatment in PSA and PSA application calculations, and especially during calculation of the risk achievement and risk decrease importance measures was performed in 2005 for SKI, the Swedish Nuclear Power Inspectorate. SKI project 14.2-041049/200402008. CCFs are dominating contributors to risk. It is therefore important with both development and correct use of CCF data. It is also important to treat CCFs in a correct and realistic manner in the calculations of the core damage frequency and when quantifying importance and sensitivity for CCF-related events, both in the basic PSA and in PSA applications. Risk increase and risk reduction measures for independent basic events (not CCF events) are calculated by setting the basic event probability to 1,0 and 0,0, respectively. This approach can also be used to calculate importance measures for CCF basic events that are created when CCF is considered. However, the interpretation of the importance measures for CCF basic events is more complex. CCF modeling usually means that CCF basic events, which really are combinations of the failures of two or more components, are introduced. In these cases, what should the probability of the different CCF basic events be during calculation of risk achievement and risk reduction measures? This paper presents results regarding risk increase and risk reduction measures for a set of example cases covering two to four trains, and with different assumptions on which CCF basic events should be set to 1.0 and 0.0. The results are discussed and are the basis for the presented recommendations, which are of importance both in calculating the basic risk importance measures, and in risk follow-up and risk monitoring applications when a component really is unavailable due to preventive or corrective maintenance, and there either is or is not knowledge of whether the unavailability is related to CCF or not. A planned activity is a no CCF case while a case without knowledge can be CCF and test of other components gives information and can be a ground for a case with complete dependence. Sometimes, an importance measures are calculated for a group of basic events, e.g., basic events for the same component type or basic events representing components in a specific physical location. In this case, it is of interest to know what attributes are assigned to the CCF basic events. Shall the original basic event attributes be assigned also to the CCF basic events? Finally, a special importance measure for CCF groups is proposed. We consider the reliability analysis of phased-mission systems with dependent common-cause failures in this paper. A phased-mission system (PMS) is a system supporting a mission characterized by multiple, consecutive, and non-overlapping phases of operation. System components may be subject to different stresses as well as different reliability requirements throughout the course of the mission. As a result, component behavior and relationships may need to be modeled differently from phase to phase when performing a system-level reliability analysis. This consideration poses unique challenges to existing analysis methods. The challenges increase when common-cause failures (CCF) are considered in the analysis. CCF are multiple dependent component failures within a system that are a direct result of a shared root cause, such as sabotage, flood, earthquake, power outage, or human errors. It has been shown by many reliability studies that CCF tend to increase a system's joint failure probabilities and thus contribute significantly to the overall unreliability of systems subject to CCF. We propose a separable phase-modular approach for the solution of fault trees, our chosen formalism, as one way to meet the above challenges in an efficient and elegant manner. Our methodology is twofold: first, we separate the effects of CCF from the base PMS analysis using the Total Probability Theorem and the concept of a common-cause event space based on elementary common-causes; next, we apply an efficient phase-modular approach to determine the reliability of the PMS. The phase-modular approach employs both combinatorial and Markov-chain solution methods as appropriate, based on the characterization of a given phase module. We use the example of a hypothetical system subject to CCF to illustrate the basic steps of our proposed methodology. Unit specific level 1 probabilistic safety assessment (PSA) models and results are available for the four VVER-440/213 type units of the Paks nuclear power plant, Hungary. The Paks PSA is updated annually within the framework of a living PSA program. The major purpose of the updates is to obtain a quantitative expression of safety impacts from plant modifications. Numerous modifications have been made in the past 15 years including design changes, procedural improvements, administrative measures, etc. Replacement of the earlier emergency operating procedures (EOPs) with new, symptom-oriented procedures is a recent modification. It is a complex task to determine the risk implications of this procedural change because the operators had to adopt a new philosophy (approach) to taking emergency responses and the operation of the control room crew as a team had to significantly change too. Efforts are made, over and above the annual PSA updates, to model and quantify the impact of EOP improvement. These efforts include review and modification of the accident sequence models in accordance with the new procedural requirements for emergency operations, re-definition of human failure events as post-initiator errors in the PSA model, quantification of operator reliability for the re-defined failure events, and re-quantification of accident sequences: • The accident sequence models have been reviewed for each initiating event of the level 1 PSA. During the review the operator responses were determined on the basis of the symptom-oriented EOPs. • Re-definition of human failure events was an embedded task within the update of the accident sequence models. The definition of operator errors was based on an initial understanding of the likely human failures that might occur when using the new procedures. • Much attention is paid to the quantification of operator errors. In the original PSA for Paks data and insights had been used from about 200 PSA driven observations at the Paks NPP training simulator. New simulator observations were made after EOP improvement to help understand the strengths and weaknesses of the control room crews and to collect information that could possibly be useful for quantitative human reliability analysis. These observations covered 4 emergency scenarios and each of the 24 control room crews at Paks. A pre-defined taxonomy was used to collect data by an observer team of 6. This was supplemented by automatic data collection to help reproduce scenario details. Simulator insights were used in combination with expert opinion to update the earlier, decision tree based quantification method. • The results of updated human reliability analysis (HRA) will be incorporated into the PSA model for the purpose of re-quantifying accident sequences. These results will include the final definitions of post-initiator human errors and the associated human error probabilities. PSA re-quantification will cover full power PSA with internal events, fires and flooding as well as low power and shutdown PSA with internal events. This paper presents the analysis steps of the Paks PSA update due to EOP changes. Also, it discusses the analysis methods, data and the most important results. Novelties in using data and insights from simulator observations are highlighted. Preliminary oil risk estimates showed that oil transportations pose a significant core damage risk for Loviisa PWR located at the southern coast of Finland. Oil transportation has increased in the Gulf of Finland during the last few years and will further increase in the coming years enhancing the importance of this risk. Therefore, a more detailed assessment of oil spill impact frequencies has been performed for Loviisa nuclear power plant (NPP) taking into account hydrometeorological conditions, oil properties and leakage sizes as well as recent oil spill statistics. During cold shutdown states the loss of service water due to excessive sea vegetation or other clogging material in the sea is an essential risk contributor because the normal residual heat removal system and the emergency cooling systems are cooled by the same service water system. Oil of tankers and cargo-carrying ships as well as other potentially fouling cargo may clog the sea water cooling systems. Nevertheless, the crude oil transportations in the Gulf of Finland pose maybe the biggest threat. In the preliminary oil risk estimate for Loviisa NPP critical amount of oil was estimated to hit the plant with a frequency of 8·10 −3 /year. This estimate was based on the frequency in the Gulf of Finland for oil spills of tens of thousands of tons and larger, which were assumed to hit the plant with certainty. The preliminary core damage frequency due to an oil spill was 5.4·10 −8 /year during power and 8.6·10 −6 at annual refueling outage. The new oil spill frequencies for the Gulf of Finland were estimated based on worldwide and Gulf of Finland oil spill statistics and seaborne crude oil transportations. In this study, oil spills ≥ 136.4 tons (equivalent to approx. 1000 barrels) were taken into consideration. A uniform oil spill probability distribution was assumed for the Gulf of Finland. Furthermore, no distinction was made between the different seasons and the probability of an oil spill. The assessment methodology of the oil impact probabilities is based on simulations of several oil spill scenarios with different intensities and durations in realistic weather and sea current conditions. Ten hypothetical spill points were selected on the main transportation route of the Gulf of Finland and five points on supplementary routes. For every spill point, different spill scenarios were studied ranging from 10 tons instantaneous spill to 144000 tons spill lasting ten days with 60 tons and 600 tons per hour spill intensities. For every spill scenario, trajectories were calculated representing space and time histories of the hypothetical spills starting from the selected spill points. Calculations were carried out for spills starting with ten minutes intervals for every day of a representative year excluding winter time when there is an ice deck on the sea. For every spill point 30240 trajectories were calculated. In the first stage of the calculations, preliminary dangerous scenarios were selected for further detailed calculations, in order to estimate the impact probabilities and average oil mass reaching the target areas. Impact probabilities were quantified for 10 μm and 50 μrn thicknesses of oil in the target areas, based on sensitivity calculations, which showed that the probabilities are negligible for oil thicknesses of more than 100 μm. As a result, probabilities of oil impact in the target areas for each spill point and spill scenario within one to ten days of the spill were obtained. In order to cover the whole Gulf of Finland and thus, include the impact of oil spills occurring in areas beyond the calculated points, 24 additional hypothetical oil spill points were added to the main transportation route utilizing the data from the in detail calculated points. The final results were obtained by combining the determined oil spill frequencies and the impact probabilities within ten days from the 39 spill points. The results show that when oil spills ≥ 136.4 tons are considered the average frequency for 10 μm thick oil to appear in the cooling water intake area is 4·10 −3 /year at power, when there is no ice deck on the sea, and slightly higher at annual refueling outage. However, for 50 μm thick oil the frequencies are over two times smaller. The estimates of the core damage frequencies seem to decrease with approximately 60% and 40% of the preliminary estimates for power and refueling outage, respectively. The effect of different variables on the result is presented in more detail in the paper. A comprehensive PSA Level 1 and Level 2 has been performed for the NPP Ringhals 2 from 2001 to 2004. After this there has been some minor updates by RAB. In this paper the Level 1 work and results will be presented. The PSA considered the operational states Power Operation, Start-up, Hot Standby Hot Shutdown. As far as possible, a complete set of initiating events was determined for the event groups LOCAs inside Containment, LOCAs outside Containment (V-LOCA), Transients, Transients initiated by Common Cause Initiators (CCI), Area events (e. g. Flooding), Fire events, Risk of Outside Core Events, External events. For each initiating event (except fire) an event tree has been constructed describing in a logical manner the accident sequences depending on the success respectively failures of the safety functions required to cope with the event in order to avoid core damage (core melt). Based on a database with the routing of cables (power, signals) from the cubicle to the respective component the cable routes were integrated in the PSA model thus taking into account all cable related fire dependencies. For all PSA relevant systems a Failure Mode and Effect Analysis (FMEA) was performed. Based on the FMEA system fault trees were constructed for each safety function in the event trees. The fault trees model explicitly independent component failures, unavailability due to maintenance/repair, dependencies on support systems, e. g. power supply, start signals and implicitly common cause failures (CCF) grouping components into a CCF group for which a parametric CCF-model is applied. The loss of the spent fuel pit cooling was identified as the outside core event that has to be analysed. The external event scenarios to be modelled were selected in a detailed external events analysis. The relevant scenarios involving single and multiple external events were included in the PSA model and quantified. Plant damage states (PDS) are the link between level 1 and level 2. Level 1 event tree sequences were grouped into particular plant damage states for which accident progression is similar. The link is performed in the same code (RiskSpectrum PSA Professional). The “small event trees — large fault trees”-methodology was applied. RiskSpectrum PSA Professional was used to model the event trees and the fault trees and to perform the calculations. LOCA frequencies were derived from the Swedish SLAP-database. Frequencies of Transients were determined using the operating experience of R2. CCI-frequencies were calculated on the base of component failure and repair rates. Fire frequencies were assessed for each fire cell using the Berry method with fire frequencies for buildings according to the Swedish X-Book. External event frequencies were estimated on the basis of site specific statistical data, e. g. extreme weather conditions. The frequencies of Area events (Steam discharge, Flooding) are determined on basis of frequencies given in the PRA Procedures Guide. The initiating event of the outside core events was defined as loss of fuel pit cooling function of the train in operation.. As far as possible plant specific component failure data from T-Book 5 were used. Where those were not available other reliability data sources or engineering judgement were used. Unavailabilities due to maintenance and repair were supplied by Ringhals based on R2 operating experience. For common cause failures the alpha factor model was used and the alpha factors mostly taken from INEL-94/0064 which contains alpha factors including uncertainties derived from US nuclear power plants operating experiences. Human reliability analyses were performed using the THERP (NUREG/CR 1278) methodology and ASEP, a simplified THERP methodology. The core damage frequency and its uncertainty is presented. The contribution of systems to the core damage frequency has been determined and is presented. A full-scope Probabilistic Safety Assessment (PSA) has been completed for the Novovoronezh Unit 5 Nuclear Power Plant (NVNPP-5), the first Russian WER-1000. The analysis has been carried out within the framework of the joint Swiss-Russian PSA project (i.e., the SWISRUS). In early 2002, the SWISRUS project entered its third and final phase, involving the extension of the PSA to non-full power modes of operation (i.e., shutdown, refueling, startup, and accidents involving the spent fuel pool) and development of various regulatory guidelines for risk-informed applications. The paper provides an overall summary of the scope, methodology, results, findings and lessons learnt from the Novovoronezh full-scope PSA study. The paper addresses the major contributors to the core/fuel damage frequency, and the insights gained with focus on the assessment of risk of severe accidents for non-full power modes of operation. Different outage types such as annual refueling and maintenance and unplanned outages are considered. The actual outage schedules and the plant records have been investigated to identify the various plant operating states and their associated duration. To enhance the completeness of the information utilized in the Shutdown Probabilistic Safety Assessment (SPSA), unplanned outages were examined. This investigation resulted in a total of 18 groups of plant operating states that were defined. The set of plant operating states modeled in the SPSA include the transition between various modes of operation. The study has covered risks originating from damage to the reactor core, fuel handling accidents and other excore accidents such as loss of spent fuel pool cooling. In evaluating risk contributors internal initiating events, internal and external hazards were considered. The SPSA addresses 62 internal initiating event groups, including heavy load drops. The initiating event frequencies have mainly been derived from WER-1000 operational experience with plant specific updates. Event tree sequences have been developed and quantified using the SAPHTRE computer code. The existing at-power fault trees have been modified to include shutdown operational configurations. The human actions that have been modeled were identified by reviewing the various plant procedures, other industry SPSAs, and through operator interviews and surveys performed at the Novovoronezh plant. Plant-specific thermal hydraulic analyses to estimate the time to core/fuel damage have been performed to support accident modeling and human reliability analysis. The full-scope PSA results are being actively used by the Novovoronezh unit 5 plant in their implementation of a risk management program to support the plant hardware and operational improvements. In 1995, the Nuclear Regulatory Commission issued a Policy Statement supporting the increased use of Probabilistic Risk Assessment (PRA or PSA) in all regulatory matters. Substantial progress has been made, but the vision of a broadly re-focused regulatory program for nuclear reactors is yet to be achieved. This paper presents thoughts on expanding and accelerating the implementation of the 1995 Commission Policy, and on how to achieve and communicate the changes that would follow for both operating reactors and new reactor designs. The result should be an enlightened regulatory approach, integrating and optimizing reactor safety, security and preparedness through the application of appropriate risk insights, using the best available information from operating experience and research, and the best available techniques, including risk-informed and performance-based regulation. The NRC can and should shift its regulatory attention, its regulatory requirements, and its regulatory practices away from riski-significant issues and to risk-significant ones. This change is desirable for the regulation of currently operating nuclear power plants, and is essential for regulation of future designs. This paper briefly reviews how UK legislation on health and safety has developed over 150 years from an initially prescriptive, industrial activity focus to a goal setting approach. The basis of current UK law is the concept that risks should be controlled and reduced so far as is reasonably practicable. Regulation of hazards requires an assessment of the risks and what measures might reduce them and implementation of those measures that are reasonably practicable, taking account of the sacrifice in doing so. The UK regulatory approach to risks is described with particular emphasis on nuclear safety and current developments on assessment of risk. Whilst quantitative risk analysis contributes to the overall judgment, the importance of the quality of engineering and safety management is also highlighted. Some pitfalls and errors that arise when using quantitative risk assessment are also noted. This paper describes the role of risk insights in the safety analysis of French Nuclear power plants. Although the regulatory position is and remains deterministic, the probabilistic approach takes an increasing place in the safety decisions. Despite a limited regulatory framework, in fact the risk insights are more and more taken into consideration as a supplement to the traditional deterministic demonstration. France holds a rather unique position with a large, highly standardized reactor fleet, build by a single constructor and operated by a single operator. This situation has provided favourable conditions for the application of PSAs to power reactors since the 80s, explaining also the rather not prescriptive regulation, which is more a continuous technical dialogue between the regulators and the operators. The presentation gives some views on the position of the French regulators (ASN), in particular on the following issues: • usefulness and limitations of PSAs • probability-based safety criteria ASN considers that PSAs are. useful tools for assessing safety, but that one must remain aware of their limitations, like scope restriction and uncertainties. Probabilistic results are always to be used as a supplement to other safety considerations, and preferably in relative terms. In order to clarify the acceptable approach, the regulators have requested the writing of a Basic Safety Rule on “development and utilisation of probabilistic safety assessments”. Regarding the second item, ASN considers that probabilistic objectives can be used as orientation values but not as regulatory limits. This is done in particular for the EPR project. Generally speaking ASN's policy is to regularly increase safety, and for that purpose Safety Objectives have not to be defined in probabilistic terms (the compliance is very difficult to demonstrate and they could have a negative effect by limiting the safety efforts when the objectives are met). This position can be illustrated by the description of some PSA applications for safety improvements. In particular for the plants in operation PSA is an important part of the periodic safety reviews and recently, for the third periodic safety review of the 900 MWe PWR series, PSA level 1, Fire PSA and PSA level 2 were considered as useful supplements to the traditional analysis. Practical examples of applications are given. The use of PSA for the safety demonstration of the future project EPR is also indicated. In 2003, the Federal agency chartered with supporting the development of improvements in healthcare safety, the Agency for Healthcare Research & Quality (AHRQ), supported the trial application of probabilistic risk assessment (PRA) in healthcare applications. One such study was the application of PRA to examine the effects of changes to the protocols used to ensure that incompatibilities did not occur between the blood types of donated organs and the organ recipients, as had happened in a widely publicized case involving a young girl who died after receiving wrong blood-type lungs and heart in February 2003. Performing the PRA involved a very extensive exploration of the processes involved in the selection and matching of organs to donors, including numerous visits to organ procurement organizations and hospitals, to observe and understand the processes involved in preventing incompatible implanting of organs, and to understand how these have changed over time, from before the event of February 2003 to completion of the study in December 2004. This study identified several features of the transplant process that are not obvious. First, the greatest ‘sin’ in transplantation is to waste a ‘good’ organ, since so few are available compared with the number of potential recipients. There is an extensive computer-based process to match available organs with already identified candidates (including confirmation of compatibility of blood types), but sometimes no match is found after an extensive search. In such a situation, a process called an ‘open offer’ can be made, where transplant centers are asked if they can use such an organ. It was found that under such conditions, several of the barriers against transplanting incompatible organs were bypassed without it being recognized as such. Several changes in response to the event in February 2003 have been made that have reduced the likelihood of an incompatible implant by about two orders of magnitude, as assessed in the PRA. PRA has yet to be embraced by the healthcare industry, which relies still on simple models of accidents—particularly single-failure models embodied in tools like FMEAs—and a belief in the ‘skill of the craft’ and personal accountability. However as this study showed, the path to failure was much more complex that could ever be captured by a tool like FMEA, and did not involve any personal failures on the part of the surgical teams (even though the head surgeon took personal responsibility for the outcome). Analyses using PRA allows the complexity of events to be identified and explored, and can show how ‘simple fixes’ like adding another blood type test immediately prior to implanting an organ can have the effect of potentially increasing the likelihood of organ failure through increasing the ischmic time of the organ—the time it is without blood flow outside of the body. However, it is recognized that, while the tools used in performing a PRA look relatively simple (as they are), it is the acquisition of the deep technical knowledge of the subject area and its representation in the PRA models that is time consuming. This study took about 4 staff-months to perform over a period of about 10 months, of which 3.5 staff months were spent understanding the processes involved and representing them in the PRA. The staff at three Washington State hospitals and Battelle Pacific Northwest Division have collaborated to apply Failure Mode Effects and Criticality Analysis (FMECA) to assess several hospital processes. The staff from Kadlec Medical Center (KMC), located in Richland, Washington; Kennewick General Hospital (KGH), located in Kennewick, Washington; and Lourdes Health Network (LHN), located in Pasco, Washington, along with staff from Battelle, which is located in Richland, Washington have been working together successfully for three years. This effort was initiated in response to the new patient safety standards implemented in July 2001 by healthcares major accrediting agency (“i.e.” Joint Commission for the accreditation of Healthcare Organization), and the hospitals' desire to be more proactive in improving patient safety. As a result of this collaboration an eight step process for applying Failure Modes Effects and Criticality Analysis (FMECA) to healthcare was developed and applied on seven high-risk medical processes. During the assessment exercises high-risk processes were explicitly defined and assessed and their weaknesses identified. Based on those insights corresponding system improvements were devised. FMECAs were performed for: • correct site surgery • prevention of patient falls • correct labeling of diagnostic images • correct medication ordering • correct delivery of solid oral medication • correct blood type transfusion • correct antibiotic IV administration Based on this collective experience, insights about the benefits of applying FMECAs to healthcare processes have been identified. And while finding suitable support for conducting an FMECA is not without its challenges, the benefits yield insights that might not easily be obtained by other means. In its 2004 report, “Patient Safety, Achieving a New Standard for Care,” the US Institute of Medicine stated that “patient safety systems [should] integrate with clinical information systems.” Recommendation one in this report was that “Improved information and data systems are needed to support efforts to make patient safety a standard of care” and that all healthcare organizations should “capture information on patient safety… and use this information to design even safer care delivery systems.” Medical incident reports typically indicate the harm (or precursor to harm), staff roles involved and potential causes related to the incident. We have hypothesized that it is also necessary to more explicitly model, and identify in (or extract from) incident reports the portions of the healthcare process that were involved in the incident; so that the parts of the process contributing to the incidents can be redesigned or modified to improve safety (e.g., reduce risk). This paper presents a conceptual framework relating incident reports to the healthcare processes and some preliminary analysis results. The results are based on incident reports collected by Johns Hopkins University Quality and Safety Research Group for a project that has established an ICU Safety Reporting System (ICUSRS) in partnership with the Society for Critical Care Medicine (SCCM) and the use of data visualization tools. Medicine is a complex, safety-critical and highly interactive system. Despite efforts to provide safe, effective care, adverse events still occur — clinicians make diagnostic and therapeutic errors, system constraints impact the coordination and delivery of care and patients suffer unexpected complications and injuries. Our current understanding of the nature of medical adverse events, and our ability to develop durable preventative or mitigating strategies has been hampered by a somewhat out-dated and inadequate model of the clinical system in which care is delivered, particularly with respect to risk and reliability. Traditional models of clinical risk have focused almost exclusively on patient factors (e.g., co-morbidities such as cardiovascular disease or diabetes), provider factors (e.g., a surgeon's prior experience with a specific procedure) or the specific clinical procedure being performed (e.g., the technical and anatomic aspects of a cardiac bypass procedure). Absent from such models are the interactions and inter-dependencies between different system components (staffing, instrumentation, protocols, procedures, access to and quality of information, communication modes, and scheduling cycles, system-wide volume and acuity, throughput pressures), an understanding of the reliability of such components under different operating conditions and the significance of failure during different phases of care. In an effort to improve the understanding of medical error and adverse events, we have been modelling a series of clinical processes using the ITEM Quantitative Risk Assessment System (iQRAS) for a major Harvard Medical School teaching institution. The iQRAS tools have proved particularly useful for modelling complex interactions between clinical subsystems and for modelling variations in risk across different phases of care (e.g., pre-operative planning phase, intra-operative phase, intensive care unit phase, recovery phase, post-discharge phase) using the mission-timeline approach. This paper describes the use of the iQRAS tools to perform a powerful risk assessment in one high risk clinical domain, and how the models are being generalized for application to a broad range of other healthcare settings. The Space Shuttle is the most capable, versatile, and reliable space-faring vehicle in the world today. As a major transportation link between Earth and Low Earth Orbit (LEO), the Shuttle has kept the United States on the cutting edge of space exploration and scientific discovery for the last two decades. It is the launch vehicle for all U.S. and many international components of the International Space Station (ISS). Loss of a Shuttle during flight has severe consequences including loss of a significant national asset, loss of national confidence and pride, and, most importantly, loss of human life. Although the Shuttle is very reliable, space travel is inherently risky. To understand the risks associated with space flight, the National Aeronautics and Space Administration (NASA) historically performed highly detailed and comprehensive qualitative risk assessments through the Critical Items List (CIL) and Hazards programs. While these qualitative programs were useful and informative, they did little, if any, to segregate and prioritize risks. With the inherent risky nature of the Space Shuttle Program (SSP), and the large number of potential hazards, the ability to quantify risks due to various hazards is also desirable. In March 2001, the SSP made the decision to develop a Shuttle Probabilistic Risk Assessment (SPRA) to help quantify and rank risks associated with the Shuttle. The purpose of this Probabilistic Risk Assessment (PRA) is to provide a useful risk management tool for the SSP to identify strengths and possible weaknesses in the Shuttle design and operation. The primary end state for this assessment is Loss of Crew and Vehicle (LOCV) and is limited to mission/flight time only. Therefore, accident scenarios initiated and realized prior to launch are not included in the SPRA. However, scenarios initiated prior to launch, but realized during flight are included. The SPRA uses a multi-phase, top down approach to identify and analyze potential risk contributors. Most ground-based facility PRAs only include events from steady state operation with redundancy and several possible mitigation paths before reaching the undesired end state. Therefore, specialized event trees are required to establish the failure event sequences. The SPRA takes a unique approach by using a single event tree entry point: launch. Assuming launch as the single event tree entry point is reasonable, because Shuttle event sequences do not generally have complicated mitigation paths and are relatively short. Therefore, the event sequences may be included in fault trees. The SPRA includes three phases of operation: ascent, orbit, and descent. Each phase represents a different mode of operation. Ascent essentially begins with T-0 (i.e., launch) and concludes with orbit insertion. The orbit phase then is assumed to continue for nine days up to reentry. The descent phase covers reentry, approach, landing, and rollout to wheel stop. As a result, the Shuttle begins as a rocket launched into space, converts into a spacecraft orbiting the Earth, which reenters Earth's atmosphere, glides to the runway, and lands. Most of the Orbiter systems are normally operating during the assumed nine-day mission, and some systems are on standby as either backup or until needed for their role in the mission. This paper discusses the scope of the SPRA and the overall methodology, shows results, and provides risk insights for this assessment. The calculated risk shows that a majority of the risk is initiated during ascent, but is realized during entry. For example, a critical hit by debris during ascent on the thermal protection surfaces does not cause LOCV until reentry. The scope, assumptions, uncertainties, and limitations of this assessment help put its results in perspective for management's decision-making process. The Space Shuttle Program (SSP) has traditionally relied upon qualitative risk assessment techniques such as Failure Modes and Effects Analysis (FMEA) [1] and Hazard Analysis (HA) [2] to identify and characterize Shuttle program technical risks. These risks comprised a safety baseline for Space Shuttle operations. It was thought that Shuttle risks could be effectively managed through careful management of the established risk controls and comprehensive pre-flight control verification actions. Additionally, the SSP relied upon a highly detailed requirements base, with extensive waiver documentation, to define and control risks related to requirements non-compliances. Overall, this approach to risk management was useful, but had significant limitations. One of the primary difficulties posed by this approach was complexity. The safety baseline for the program at any given time was represented by thousands of hazard causes, critical items, and program waivers — making it difficult to understand the level of risk that flight operations posed. It was also difficult to understand the integrated effect of proposed changes to the shuttle systems, operations, and procedures. Efforts to prioritize safety risks were confounded by this risk management model as well, since it was difficult to compare competing failure modes, hazards, and requirements non-conformances based on the potential risk posed through hazards, failure modes, and waivers. The SSP began developing PRA models during the 1980s on a piecemeal basis, but did not attempt a system-level PRA until 1995. This landmark study made some important conclusions regarding Shuttle risk that have remained true throughout later assessments. However, the analysis included only some of the Shuttle's critical systems and operations, and did not include significant risks associated with orbit phase operations. Subsequent studies were conducted to gradually increase the completeness and fidelity of the PRA. While these efforts suffered from a lack of consistent standards, limited program sponsorship, and limited funding; additional progress was made in developing a system-level PRA. The latest iteration of this activity, initiated in 2003, has enjoyed a greater level of support and sponsorship from within the program than any previous analysis and has resulted in a higher fidelity, peer-reviewed study, reflecting all three unique mission phases: Ascent, Orbit, and Landing. While additional areas remain for improvement in the SPRA, the current iteration has provided the SSP with a system-level model that can be used to support decision making. Additionally, smaller scope, focused PRAs have been spun off to address specific technical decisions, and for the first time, probabilistic analysis is being used to resolve technical issues, define flight rationale, and inform real time decisions during Shuttle missions. This paper describes the development of the SPRA and how the SSP has adapted its risk management paradigm to include PRA. It describes how the previous SSP safety risk management paradigm has evolved and how PRA is being used to influence program decisions. Lastly, this paper proposes a model for integrating PRA into program risk management and identifies areas that need further improvement in SSP practice. The Columbia Accident Investigation Board (CAIB) report had provided NASA with a set of comprehensive recommendations to improve the safety of the Space Shuttle Program. At the same time, NASA also pursued an Agency-wide effort to take a fresh look at all aspects of the Space Shuttle Program, from technical requirements to management processes, and developed a set of internally generated actions that complement the CAIB recommendations. The resulting implementation plan is the path that NASA took to safely return to flight. One “return to flight” technical issue is the potential hazard of an uncommanded reaction control system (RCS) firing while the Orbiter is docked at the International Space Station (ISS). This paper shows a quantification of the likelihood of an uncommanded RCS firing because of wire failures while the Orbiter is docked at the ISS. The fault tree methodology is based on known failure modes and operational considerations that potentially can lead to damage. The failure modes in the model are connector pin-to-pin shorts, exposed wires due to vibration damage (chafing), heat damaged wires, arc faults (undetermined cause), resistance heating from degraded splices, undetected maintenance damage, and operational damage due to other operating causes like delamination, exposure to chemicals and elemental oxygen. Quantification based on assumptions and limited anomaly data resulted in 1.4E-4 failures per ISS docking mission. The failure is judged to be an improbable event that is much smaller than the failure probabilities of other hazards to the Orbiter. Common cause failures are dominant as they affect multiple wires simultaneously. The two most likely failure modes are arc fault and chafing. The results of this study aided in the management of ‘return to flight’ risk by helping prioritize proposed RCS wiring upgrades with respect to other ‘return to flight’ safety issues. We developed a prototype of human-in-the-loop simulator of inter- and intra-agency communication and coordination in disaster response. The base simulation is a multi-agent simulation of emergency response in which each organization (group, section, or organization) is implemented one cognitive agent and thus simulation of interaction between different organizations with the same such granularity is possible. We constructed the simulator by adding user consoles for entering and receiving messages from these computer agents via CORBA platform (a distributed object computing environment). This user console is expected to allow a user to use the four types of messages: inform, query, request, and resource. These message types along with the ontology are designed to be common between the agents and human users. Still some efforts are necessary to enrich the interactions between the agents and humans, this kind of simulation could contribute to understand the influences of actual human decisions and commands to the entire performance of emergency response system. In Europe, an increasing number of road tunnels are realized. In 2002, 398 Trans European Network (TEN) tunnels existed, and between 2002 and 2010, additional 114 TEN road tunnels are realized [1]. Several of these tunnels cross national borders, such as the Mont Blanc (France-Italy) or the Fréjus (Spain-France). In particular the cross-border or trans-boundary aspect might hamper the emergency response actions in tunnels in case of an accident for example caused by language problems. Examples of these hampered actions concern the Mont Blanc tunnel inferno and the Eurotunnel fire. These accidents made clear that cross-border tunnels have peculiar aspects that make the rescue of people in the tunnel extra difficult including: • language: are emergency responders from different countries able to communicate to each other? in which language are the tunnel-users informed and in which language are the contingency plans written? • coordination: who communicates with whom? • hierarchy: who is in charge in the emergency response operations? • equipment: can equipment of one country's emergency responders be used at the other country's part of the tunnel? • preparation: which scenarios form the base for emergency response preparation and how do the emergency responders conduct their training exercises? The EU recognized such issues and aimed at harmonizing the various procedures and appointments in the EU countries [1]. The EU stated: “ Insufficient co-ordination has been identified as a contributory factor to accidents in trans-boundary tunnels. Moreover, recent accidents show that non-native users are at greater risk of becoming a victim in an accident, due to the lack of harmonisation of safety information, communication and equipment ”. To harmonize the cross-border aspects of tunnel emergency response actions, and part of a broader European Union program, called SafeT, Nibra studied typical cross-border intervention aspects and formulated guidelines for harmonization. Literature study and case studies of the Mont Blanc and the Eurotunnel revealed the elementary cross-border aspects that hampered tunnel emergency response actions. A study of operational tunnel procedures of the Fréjus (France and Spain), Chunnel (France and United Kingdom), Öresund (Denmark and Sweden) and Karawanken (Austria and Slovenia) revealed multiple solutions to deal with various cross-border aspects. In the end, fourteen guidelines resulted. Because the guidelines not only concern language issues but also coordination, organization, training and liability issues, such guidelines can be useful for all tunnels crossing any administrative border where multiple emergency response organizations are responsible for fighting accident consequences. In life-threatening emergencies, citizens need to be warned. The currently used method for citizens warning in The Netherlands is the use of a siren. Modern telecommunication technologies have great potential for informing the public. Especially the use of text-based features of mobile phones is considered for this function. Advantages of such a system are that these technologies overcome problems of hearing the siren and that text-based messages provide possibilities for giving more and more detailed information. In a number of pilot studies, Delft University of Technology has gained experience with the possibilities of using mobile phone technologies for citizens warning services. Up to this moment these studies have focused on the technological potential in terms of the percentage of population which can be reached, some technological deficiencies of the systems in development as well as acceptance of the new warning system by the public. However, an effective introduction of a new warning system requires more than overcoming the technological challenges. This paper discusses research into what information should be given to citizens in case of an emergency. What kind of information do people want to receive in an emergency and what kind of messages are understood best by citizens are the questions that will be answered in research that is being carried out at this moment by the Safety Science Group of Delft University of Technology. This paper will describe the method of research, some preliminary results and some indications for further research that will be carried out in the first half of 2006. This research described in this paper is part of a number of large scale pilots in which the use of cell broadcast is studied in the Netherlands. The Safety Science Group is involved

The 5th International Conference on Computer Research and Development (ICCRD 2013) was held in Ho Chi Minh City, Vietnam, February 23-24, 2013. The ICCRD 2013 was hosted by the Science and Engineering Instititute. The main objective of ICCRD 2013 was to provide a platform for researchers, engineers, academicians as well as industrial professionals from all over the world to present their research results and development activities in Computer Science. This conference provides opportunities for the delegates to exchange new ideas and application experiences face to face, to establish business or research relations and to find global partners for future collaboration. This front matter contains the preface and welcome message, committees, and the contents. This paper mainly presentsthe multi-objective coordination evolutionary game processwith complex network models. It is found through the experiment that the average payoff and the average vector matching degree of the nodes tend to be stable and the nodes are clustered with the evolution of the network model. Through the structure of the network and the detailed distribution of vector matching degree, some interesting phenomena which reflect the real life are found. Recently, social networks have been developed rapidly, especially for Facebook which is very popular with 10 billion users. In this paper, we analyze the relationships in social network from the perspective of complex networks characteristics, comparing with BA scale-free network model, WS small world network model and ER random network model. It is indicated in the experiment that Facebook social networks don’t completely satisfya certain existing complex networks model, butit is similar tosome kind of network model in different complex networks properties. The results reflect the existence of the community structure in the social network, so the node degree is not a simple inverse relationship with average path length like the normal three network models, but varies with the location distribution of the node in the community. To study the evolutionary characteristics of public opinion spread in social networks,a new propagation model is proposedbased on epidemiological model and accords with popular online network spreading. According to this propagation model, a large number of simulation experiments considered the influence of users’ different characteristics are finished. The results ofsimulations showsome evolutionary characteristics— public opinion outbreaks in early stage of spreading, influence range of news reducesprogressively and periodically in middle stage and public opinion presents dynamic stabilityin late stage. According to game theory, a mechanism of altruistic behavior based on reputation and future expectation is proposed in this paper. In this mechanism, “Reputation” is used to evaluate the historical behavior of one player and predict the probability of the player choose cooperation. Players tend to choose the one with higher reputation as their opponent. The total payoff consists of current payoff and futurepayoff. Analysis and simulations show players choose cooperation voluntarily for the purpose of payoff-maximizing. The reputation oscillates with a particular mode in a specific domain of convergence. The higher the future expectation is, the greater the possibility of cooperation emergence will be. The level of convergence average reputation of the group relies on the level of initial average reputation of the group. In addition, the non-subjective factors exist in the real world a resimulated. A slight increase of the non-subjective factors can effectively reduce largely the average convergence reputation of the group. These factors seriously prevent the cooperation from emerging. The multi-domain wireless sensor networks are a set of single network co-exist in the same physical location by different authorities. Since nodes are often battery-limited, the important design criterion is the maximization of network lifetime and the efficiency of balancing energy. In this paper, we present a modified topology control protocol. And a new algorithm BiPEL which use credit incentive is proposed. The results show that the BiPEL algorithm improves the network lifetime and balances network energy more efficiently. The lifetime of networks is not positive correlation with cooperation rate. Firstly, this paper presents the concepts and key technologies of business intelligence. Then three aspects have been described in detail, such as constructing a data warehouse for multidimensional data analysis, the establishment of OLAP multidimensional data analysis and data mining based intelligent technology to intelligent community. We can analyze the user behaviors by different methods, give them reasonable recommendations and guide them to scientific rational use of electricity. Deficit of classified texts for different languages especially Persian language is one of the main problems in developing text information retrieval and statistical machine translation systems. A new method of text classification in Persian language will be introduced in this paper. For this purpose, in the first step the keywords of different parts of texts will be extracted employing the classified documents of Wikipedia and through a probabilistic method. Then by employing two methods of machine learning which are based on k_Nearest Neighbor and decision tree algorithms, the texts will be classified in their specialized fields. The evaluation results reveal that this method is appropriate for classification of Persian texts. The results also show that the nearest neighbor method is more applicable to classify texts in comparison with decision tree method. Memetic Algorithms (MAs) attempt to improve evolutionary process by incorporating problem-specific knowledge into their framework. Consequently, MAs may deal well with the highly constrained combinatorial problem such as timetabling. Memetic timetabling concerns with applying MAs for solving timetabling problems. Knowledge from the problem that is included in the chosen representation, search operators or evaluation function may improve the performance of memetic timetabling. In this work, two versions of direct representations are then implemented in an MA to solve the benchmark timetabling instances. The computation results of MA using both representations are compared to measure the merit of each representation. Comparison with best-known results from literature is also shown to further look into the possible improvements that can be made on the chosen chromosome representation. In this paper, we present a new system to track eye gaze using a single camera and single light source. In system implementation, an accurate and robust eye gaze tracking eigenvalue extraction algorithm based on image analysis was proposed. The system successfully positions glint center and pupil center at sub-pixel level, and it remains a relatively high accuracy and speed in different illumination situations. Experiments show the average variance of the system is approximately 21 pixels. The method has increased the accuracy level of single-camera eye gaze tracking system, and offered a both efficient and low-cost solution for eye -gaze tracking oriented human-computer interaction. The re-infection of individuals who have successfully recovered from a specific disease is a major health concern. In the absence of vaccines and other treatments, reinfection can occur when there is movement between populations and within populations. There is a growing interest as to how to model population movements in order to find optimal control strategies. The approach proposed here uses cellular automata to model reinfection under one possible migration scheme and demonstrates how population movements can lead to different patterns of reinfection among four populations of different size. This article presents an approach to designing an adaptive, data dependent, committee of models applied to prediction of several financial attributes for assessing company’s future performance. A self-organizing map (SOM) used for data mapping and analysis enables building committees, which are specific (committee size and aggregation weights) for each SOM node. The number of basic models aggregated into a committee and the aggregation weights depend on accuracy of basic models and their ability to generalize in the vicinity of the SOM node. The proposed technique led to a statistically significant increase in prediction accuracy if compared to other types of committees. Football is one of the most famous sports in the world and also very popular in Thailand. As a number of fans are continuously growing up each year, this sport becomes a big part of business volume in Thailand. It is inevitable to have a fair and attractive tournament. This scheduling problem becomes so complex to meet various demands from associations, clubs, media and fans. Fuzzy C-means (FCM) clustering algorithm is a typical representative of the fuzzy clustering methods which are based on the objective functions in clustering algorithms. So far, a lot of literatures still have been studying the fuzzy clustering algorithms which are based on the objective functions. In order to understand these clustering algorithms and their characteristics more systematically and deeply, this paper reviews and evaluates advantages and disadvantages of each typical fuzzy C-means algorithm from three aspects: the modification of the objective function, the modification of degree of membership constraints and clustering problems of different types of data, and summarizes the prevalent problems among the improved fuzzy C-means clustering algorithms improved fuzzy C-means clustering algorithms and prospects of them. Firefly algorithm is a swarm-based algorithm that can be used for solving optimization problems. In this paper, the methodology of fuzzy c-means algorithm is incorporated into the original firefly algorithm to improve the performance. Six popular benchmark data sets from the UCI machine learning repository are used to test this technique.The experimental results showed that the proposed algorithm surpasses both the base-line technique k-means clustering and original fuzzy c-mean. Plants are an indispensable part of our ecosystem and the dwindling number of plant varieties is a serious concern. To conserve plants, their rapid identification by botanists is a must, thus a tool is needed which could identify plants using easily available information, e.g. a photograph of the leaf. This paper tries to fulfill this requirement. An algorithm to recognize and classify a digital image of a leaf is conceptualized and implemented in MATLAB; the leaf database is constructed using MySQL. The developed software uses features based on the shape of the leaf such as aspect ratio, compactness, centroid and edge measure. These features are automatically extracted from a digital image, then converted to numerical parameters to compare with the data stored in the database and the results are displayed. The accuracy of the software was tested with a specimen dataset of 45 leaves and was found to be 83%. Production factories in which stable voltage is critical, e.g., electro-plating factory, require constantly stable voltage to minimize loss by adjusting incoming voltage in real time, even if low-quality electricity is supplied from outside. To solve such problem often being raised from the factories located in the area with unstable electricity supply, we designed N-M switching control system and verified its correctness using LTL model checking technique. Granular computing has been applied in many fields to solve problems and describe many spaces at different granularity and hierarchies. This paper proposes a rule generation approach based on granular computing in the frame of rough mereology. The proposed approaches generates a single rule granule from granular space in each step instead of selecting a suitable attribute according to some measure and improves the performance of rule granule generation. Rules are generated by using voting strategy based on weights computed by means of rough inclusions induced from residual implications of continuous t-norm. Experimental results obtained show that the overall accuracy rate average of dataset is higher than other approaches applying decision tree algorithms. Also, the resulted average rule length is less than average rule length generated using decision tree algorithms. In this paper we propose a fuzzy rule generation approach based on granular computing using rough mereology (FRGAGCRM). The proposed system works in two phases. In the first phase, the pre-processing phase which use fuzzification methodology which map the numeric dataset into categorical dataset according to membership function described in this paper. In other hand, the second phase consists of the rough mereology phase and rule generation phase. The rules are generated by rough mereology and rough inclusion techniques. The experimental results using Irvine, ML Repository dataset show that FRGAGCRM is better than RGAGCRM, ID3, and RGAGC. The performance measure of FRGAGCRM is compared with the performance of RGAGCRM, ID3, and RGAGC in same dataset. The result obtained empirically demonstrates that the proposed FRGAGCRM performs superior. Separating codes have their applications in collusion-secure fingerprinting for generic digital data, while they are also related to the other structures including hash family, intersection code and group testing. In this paper, we study upper bounds for separating codes. First, some new upper bound for restricted separating codes is proposed. Finally, we illustrate that the Upper Bound Conjecture for separating Reed-Solomon codes inherited from Silverberg’s question holds true for almost all Reed-Solomon codes. The number of Twitter users in Indonesia are increasing rapidly. They sent various contents, like opinions, news, promos and insulting about some products. Highlights of sentiment analysis is to explore public opinion about 3 Indonesian brands telecommunication products. The aim of this research is to find information from Twitter content that intended to facilitate a company or owner of the products in getting the assessment of their products easily. Hopefully, the opinions can give some input for the company to improve their services and product performances. The approach method is by labeling positive, negative, or neutral on Twitter contents which have been collected. The contents of Twitter limited based on the existence of Twitter account which is in Indonesia. This research is using Naïve Bayes Classifier algorithm. Twitter content can automatically classified and gives score 80.6% for the accuration. In this study, determining damaged regions on a print head of thermal transfer printer by image processing methods is presented. The aim of the developed codes is to determine the hurt regions on resistive parts of print heads which are heated during the printing action. Photos which are used in the study are taken by a microscope and the images of resistive parts are compared pixel by pixel. Determining degradations by image processing is useful to establish the reasons of damages rapidly and it can be used for developing heavy duty, long life thermal transfer printers. Also this study can be used in some other studies of graduate and undergraduate students who are interested in areas of determining some problems of thermal transfer printers and solving them. SQL injection is serious threat for web applications these days because the injectors are well-advance in the technology and there loopholes in the web application & browsers. SQL injection may allow attackers to get unrestricted access to database. The injection will result to leak of sensitive information to the attacker and the attacker use this information for their purpose why they hacked for. There are many proposals from researchers but they are somehow failed or failed to address the full scope of the SQL injection. There are many reasons why researches proposed solution for subset of SQL injection. Implementing the full known SQL injection may block the use of SQL quires in the application. Another one with the technology attackers using new methods of injection so there is a need to update the proposal. With all these I am proposing a new idea which works fine for known injections and easy updatable for future. This is mostly conceptual paper which is to get suggestion from the experts on this area to implement most right algorithm and pattern matching for SQL injection. Attack graphs used to compute possible attack paths and hidden activities for given system configuration with information available in Anti-Forensics Vulnerability Database. The purpose of this attack graphs is to help forensics investigator to identify potential attack path(s) with possible vulnerability sequence for the attack. After attack, investigator will find out the right path with the attack graph and the configuration because the graphs will have all possible attack sequence for the attack. But there are possibilities that the evidence can be obfuscate, encrypt, minimized or eliminated if anti-forensics tool is being used. For the anti-forensics tool there are proposal to use another single node (Anti-forensics Activity) every activity but my proposal is to use Hidden-Path which has anti-forensics node in it. With my proposal not only we can identify the use of anti-forensics tool but also generate potential attack sequence. Efficient and effective energy management is a prominent factor in the new global economy. The Malaysian government is fulfilling this ambition by implementing a Nuclear Power Program (NPP) to provide sustainable energy for the country. However, the safety issue has become important in the light of public concern. The focus of this paper is to examine requirements for Malaysian radiation and nuclear (RN) contingency plan such as emergencies, incidents or disasters, originating from radiological and nuclear resources. The prospective emergency plan requirements are determined by Work System Theory (WST) modeling. A case study explanatory structural analysis is depicted as proposed network diagram and work system snapshot. The development of these snapshots can help to enhance the proposed emergency plan simulation modeling in order to achieve certain preparedness regarding nuclear safety environment in Malaysia, far better than the use of manuals or conventional table top emergency drills. This paper focuses on the research of collision avoidance local path planning for mobile robots, which is to find a collision-free path from the source position to the destination position among unknown obstacles under the guidance of the global path planning. The paper researches into the local path planning problem in the environment which the shape and location of obstacles is unknown, and proposes a new path planning method of collision avoidance. It realizes the collision avoidance by the path planning method likes the artificial potential energy field based on the grid-type obstacle environment. The method has a low complexity and high efficiency, and results out a valid collision-free path. In this paper, the details of the algorithm are introduced. Student difficulty in understanding basic concepts and scientific knowledge in new trends of information technology (NTIT) is a major obstacle when achieving teaching objectives. Abstract concepts and knowledge, which can not be seen or touched, are important issues. Teaching strategies using media (videos) can improve the learning process. This research study shows the efficiency of using videos in teaching NTIT at university, presents the problems faced by following such strategies, and shows how the cloud computing technology can help overcoming such problems. As a practical case, we present how we benefit from cloud computing to store high-capacity videos which are used in teaching NTIT in the Management of Information Systems Department at Sharjah University, UAE. Small-scale details are visually important for smoke animation but often get lost due to numerical diffusion in grid-based simulator. This paper introduces a novel interpolation technique to keep high frequency details in smoke animation. We construct a FFT-based ideal interpolator for fluid fields. We then apply it in the Semi- Lagrangian advection step of a stable fluid solver. The interpolator transforms velocity field into Fourier domain and upsample it from the base grid to a finer grid. This interpolator is ideal for reconstructing high frequency sub-grid details calculated on the base grid. We also provide effective techniques for smoothing boundaries to mitigate the ringing artifacts brought by Fourier transform. Our method is simple and easy to implement. Results show that our approach can preserve more compelling high-frequency details in smoke animation with little computational cost. Fuzzy Logic (FL) is a multi-valued logic, that allows intermediate values to be defined between conventional evaluations like true/false, yes/no, 0/1 etc. Fuzzy based decision making process that uses rules based on decision maker’s experience is applied to automate decision making system. The implementation of vendor system using fuzzy logic technology has the capability of mimicking human intelligence for choosing the best vendor when buying notebook computer on the online. As the system is implemented for a notebook computer buyer, the basic criteria typically utilized for this purpose are the capacity of CPU, memory (RAM), hard disk (HDD), price (Kyat) and warranty duration of a notebook computer. The buyer establishes a set of evaluation criteria that can be used to compare potential sources when a vendor is selected via web page on the browser. This system can be produced the suitable vendor. This system is implemented by C#.net programming language. The existence of risk in electronic transaction makes consumer reluctance to engage with E-Commerce (EC). Yet the inferior position of consumer that often looking for seller’s trustworthiness makes this kind of consumer trust unique and has gained number of researches in order to achieve deeper understand the concept of trust. Although there are various researches conducted focusing on different properties of trust but lack of attention has been given to interpersonal based trust that consider the importance of social presence (SP) in EC environment. Taking the advantage of rapid growth of social networking now days, this study has its main objective which is improving current EC implementation through SP factors. In order to achieve this goal, this study has explore all trust properties pointed by previous researches and incorporated SP factors and integrating it with existence business-to-consumer (B2C) EC trust model. It will also propose a new model that shows all tested SP factors and providing guidelines for web merchants in building a trusted B2C EC website in the future. In the field of Information security, it is a significant issue to implement risk assessment upon information systems. The ultimate goal of such action is to guide the decision makers to find a balance between “investment cost” and “security level”, so as to establish protection strategies and mitigation plans for classified asset risks. As with the development of information security risk assessment methods, security assessment experts have gradually started concentrate on the suitable risk issue, which includes highest, lowest and preferred risk. A big issue demanding prompt solution is that how to find the suitable risk from massive risk evaluating data quickly and efficiently, so as to bring it into effect upon enterprise management and decision making. In this paper, on the basis of in-depth comparison among existing multi-objective optimization techniques, Skyline and Top-k query methods are implemented in the information security assessment system to deal with the suitable risk issue in the assessment data. According to the experiment results based on evaluation data sets with different distributions, Skyline and Top-k methods are able to efficiently conduct multi-objective decision making in risk assessment, and can also be implemented in the risk assessment systems to deal with the suitable risk assessment issue. Current military information systems have been changing toward intelligent military information systems and constructing the military ontology plays a central role in this change. Using the military ontology, machines can understand the information and automatically manage it. In this paper, to show the possible application of the Mixed Ontology Building Methodology (MOBM) which was proposed in our previous research, we present the process of the military ontology construction by applying the MOBM to Army Tactical Command Information System (ATCIS). As a result, we found that the MOBM is very practical and applicable to the military ontology construction, and hence it would be widely used in the army in the future. A k -ary tree is a rootedand ordered tree such that every internal node has exactly k children. A natural representation of k -ary trees is the use of x-sequences or z-sequences introduced by Zaks in 1980. Under such representations, a loopless algorithm that generates z-sequences of k -ary trees in Gray-code order was developed by van Baronaigien (2000) and Xiang et al. (2000), respectively. In this paper, we present a new loopless algorithm to generate z-sequences of k -ary tress in Gray-code order. The development of our algorithm is in need of less memory space and is more efficient than the existing algorithms. Moreover, with a slight modification, our algorithm can generate x-sequences of k -ary trees such that only one pair of positions have different binary codes for two successive sequences. A query-answering problem is concerned with finding the set of all ground instances of a given atomic formula that are logical consequences of a given logical formula. Meaning-preserving Skolemization is essential for development of a correct and efficient method for solving query-answering problems on first-order logic. It requires global existential quantifications of function variables, which in turn require an extended space of first-order logical formulas. Instantiations of function variables in such an extended space, however, often result in generation of infinitely many models, causing several computational difficulties. This paper proposes a method for eliminating function variables from a set of extended clauses, which provides an important step towards solving query-answering problems that include unrestricted use of universal and existential quantifications. The study focuses on developing a mobile application suited for Windows Mobile Phones recognizing Philippine Herbal Leaf Images which utilizes Artificial Neural Networks as its main algorithm. The study aims to know the degree of accuracy given by the parameters such as resolution of the image, orientation of the image, illumination of the image and the distance of the image to the mobile phone. After the experimentation, it was observed that the application reached the set accuracy of 85% for recognizing leaves with the computed accuracy rate (AR) of 85.21% which is considered as acceptable. Sequential processing has dominated the computer world. Substantial speedups can be made using parallel architecture. However, parallel algorithms often require shared memory. This paper introduces a four-port Parallel Random Access Memory (P-RAM) integrated circuit emulated on a Cyclone II FPGA. The P-RAM was simulated and connected to the STORM System-on-Chip. This implementation is limited to 4kb of memory as it is optimized for speed and sacrifice flip-flops and multiplexers. Contributions are made in the design of a P-RAM and FPGA implementation and operations. These findings suggest that parallel RAM integrated circuit is viable and can be integrated into modern computer architecture. Home Automation for alarms systems is a desirable technology. There have been a number of attempts of Home Automation system. This paper presents the development of a new Home Alarm Automation System. The goal of the project is to build a new alarm system which is specifically engineered to monitor the in-door environment and warn the owner by sending an SMS to their mobile phones about any abnormal activities. Smoke, high temperature, toxic gas and unexpected intrusion are all monitored by this system. This study implements input abstraction techniques such as Natural Language Understanding with sub-processes of tokenizing, part-of-speech tagging, and parsing. The developed system called Philippine Land Law Expert (PHILEX) chatbot focused on giving answers to the users who are in need of solutions to their problem or assistance in regards to property land laws and rights. It can be used and of great help to the lawyers in assisting their clients. In addition, it can be also utilized by those who seek counseling regarding property land laws and land rights but is not intended in substituting the professional practitioners. Copula gives a flexible tool for understanding and describing the possibly nonlinear dependence structure among random variables. Copulas make a link between multivariate joint distributions and univariate marginal distributions. In the quantitative risk management, the dependence relation among risk factors is an important issue, where copulas are well employed as a convenient method. Here we deal with a family of generalized Archimedean (GA) copulas and consider the tail dependence of such copulas. The tail dependence represents a kind of asymmetry in the modelling, where copula provides a handy way of computation. We calculate the tail dependence coefficient of several examples of GA copulas and discuss its meaning. Accurate indoor geolocation is an important and novel emerging technology for commercial, public safety, and military applications. Recently, there is increasing interest in WIFI positioning systems due to the cost and availability of this technology. However, the main problem in WIFI-based localization is the severe fluctuation of received signal strength ( RSS ) even for a static client. In this work, we consider introducing a wavelet denoising filter as pre-processing block in order to remove the undesirable noise generated by radio propagation in the environment and then a radial basis neural network is used to estimate the position by using RSS . This article present the design and implementation of a Energy-Efficient Arithmetic and Logic Unit that has been developed as part of low power processor design in the platform Xilinx ISE 14.2 and implemented on Virtex-6 XC6VLX240T-1FFG1156 FPGA. Any of control signal Clock Enable, Gate Enable, Write Enable, Set/Reset, Preset/Clear, Clock/Gate uses different number of on chip logic, signal, IOs block hence never consume uniform power. Using reset for each design means we are using two reset one local and other global always increase power consumption than design with only global reset. In mapping, behavioral HDL coding reduces the number of on chip logic component utilization in compare to data flow HDL coding that ultimately reduce dynamic power consumption of our ALU Design. The objectives of this research are to determine the phylogenetic relationships among populations of Macaca fascicularis in Malaysia. This research involved 14 individuals of Macaca with locus Cytochrome C Oxidase subunit I (COI). DNA was extracted from blood, tissue and liver. Nine individuals which are species M. fascicularis, M. nemestrina, Trachypi thecus cristatus and Trachypithecus obscurus were used to generate phylogenetic relationships among populations of M. fascicularis . COI region sequences of mtDNA have been obtained in size of 700 bp from PCR process. Then, by using the genetic softwares which are Bioedit, MEGA, ClustalW, PAUP and MrBayes, the samples were aligned and analyzed. The phylogeny tree topology was constructed using the Neighbor-Joining, and Maximum Parsimony methods. The phylogeny tree topology shows that M. fascicularis do not segregates according to the theoretical phyleogeography in Malaysia. This is because the sample from Negeri Sembilan has merged together into the samples from Perak and Penang in one clade. However, M. fascicularis has successfully separated from the other species and genus thus form monophyletic group. The availability is one of the most vital prerequisite issues in providing IaaS Cloud Computing services. It means any fault must be identified and reacted reasonably in order to ensure quality of service [3]. To achieve this requirement, fault tolerance system in IaaS has to balance accuracy in fault detection with reaction rate. In this paper, we present a new approach based on fuzzy logic to detect faults happening in IaaS accurately and flexibly. Based on the study of mobile application development technology and J2ME architecture for mobile platform, the paper aims at analyzing mobile game development principle and implement method on J2EE platform by designing a Bomb Man game. It not only gives the concrete design and implement processes, but also draws a conclusion about how to design in mobile environment using object-oriented programming.

This timely volume addresses the critical issue of safe design of mechanical structures, systems and components belonging to hazardous facilities, in order to withstand the effects of extreme loads. These extreme loads may be either man-made, such as design-basis accidents, aircraft crashes or malevolent vehicle intrusions, or due to natural disasters, such as earthquakes or tornadoes. Hazardous facilities include the nuclear, petrochemical and biomedical industries. Detailed information on government regulations and industry standards is provided. The structures, distribution systems, and components addressed by this guide include those addressed by ASME and related standards, such as API, IBC, ASCE, and others. This front matter contains the acknowledgement, table of contents and forward This book has been prepared as a guide to the structural and leak-tight integrity design and analyses of structures, distribution systems and components (SSC), which are addressed by construction standards prepared by the American Society of Mechanical Engineers, ASME and other closely related Standards. Particular application is to SSC, which are subject to extreme load phenomena, whether of a design basis external natural, man-induced or accident phenomena. The book is intended for use in assisting structural-mechanical engineers in their supply of engineering services to the nuclear, petrochemical and biomedical industries. Probability of failure or risk-informed assessments are playing an increasingly significant role in defining design load for hazardous facilities [1, 2]. One important role has been in the area of the decision process regarding extreme load design bases. Prior to its employment, hazardous facility design requirements were based, to a considerable degree, on the “minimax” decision rule, which says the worst possible future is sure to be found, so action should be taken to minimize the maximum possible loss. This is popularly referred to as the “what if” basis for design, where loads and particularly their combination, which could be postulated regardless of their probability or conditional probability of occurrence, became design bases. The minimax rule was tempered somewhat by the reasonable decision rule, which says that any reasonable decision maker under the same circumstances and with the same background would take the same action; hence these decisions tended to be made taking historical precedents into consideration. Finally, there is the acceptable risk decision rule, which say: Take the action where the sum of the products of the probability of exceedence and the consequences of exceedence are equal to or less than a risk associated with some natural phenomena over which society apparently has no control, or is less than some man-made activity that historically has been accepted by society, considering the actual or perceived cost∕benefit to society. The current tendency in hazardous facility design is a continuing shift away from the minimax rule or deterministic-based siting and design toward an acceptable risk rule tempered with the reasonableness rule. This is typically referred to as risk-informed siting and design. However, it should be understood that there is often an inability to develop an accurate probability assessment in many areas because of the lack of the necessary statistical data or ignorance as to the governing physical relationships and variability in the phenomena being investigated. For this reason deterministic bounds are often placed on the applicable design criteria that are developed based on the acceptable risk rules. Use of probabilities, as a basis of design commonly referred to as risk-based design, must await the development of probability estimates with less uncertainty than is now the case. Extreme loads applied to mechanical SSC often involve time-dependent structure fluid interaction. Over the past 5 to 10 years enormous strides have been made in the ability to more accurately define these loads in the design of mechanical components. Primarily due to the strides made in the development of finite-element and difference algorithms and resultant computer programs, these loads particularly associated with system transient time-dependent loads at charges in directions and areas can be determined with an accuracy not previously achieved. To date these analyses have generally been restricted to elastic or linear response systems. However, it is expected that they will rapidly become available for inelastic nonliner system analysis more characteristic of mechanical systems at or near failure. Unfortunately, these new analytical techniques tend to be relatively expensive both with respect to the cost of the computer software and the cost of inputting and evaluating the output results. In addition, it tends to be difficult to evaluate the validity of the results. This chapter provides approximate first-order load definitions for various types of extreme loads typically applicable to mechanical SSC. These design inputs may be used where more exact solutions are not required, or can be used to independently verify more exact solutions. 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. The primary purpose of the seismic design of high-hazard facilities is to construct them so that there is no excessive exposure as a result of radiation or toxic substance release to the public and collocated workers. A major earthquake in relatively close proximity to the site may cause major damage to conventionally designed industrial facilities. For this purpose, a more stringent seismic design is usually required than would be required for conventional facilities to minimize potential for loss of structural and leak-tight integrity. For conventional facilities, in some cases ensuring the operability and life safety of its occupants is usually the primary concern, while for high-hazard facilities the confinement of radiological or toxic substances is the primary concern. The structural loads associated with extreme straight and hurricane wind velocities have usually been considered statically applied and distributed in accordance with the American Society of Civil Engineers (ASCE) 7-05 Standard for buildings and other structures to include chimneys, tanks and rooftop equipment [1] or ACI-307 [2] for concrete stacks and STS-1-2000 [3] for steel stacks. Tornado wind effects as a design-basis load have generally not been defined or required for consideration in design as a normal or severe load as defined in Chapter 3. They have been prescribed as an extreme load for nuclear safety-related structures. In some cases they have also been considered for petrochemical facilities in tornado-prone regions, but this has not become a standardized design practice associated with hazardous, petrochemical or biomedical facilities. As a result a tornado wind field and the methods used to combine the effect of wind or “dynamic pressure” and the associated reduced atmospheric pressure to include building openings and potential airborne missile effects have not been defined in general industry standards for design purposes, except as applied to the nuclear industry. A rational yet simple combination of the wind field and reduced atmospheric pressure effects for use in analysis and design or evaluation of structures subject to design-basis tornado loadings is presented in this chapter. There is also a suggested procedure for including the effect of postulated, straight, hurricane and tornado-borne missiles in structural design in this chapter as well as in Chapter 8. This chapter is intended to provide side-on and reflected static pressure loads on structures, systems and components (SSC) from an external explosive blast. These dynamic pressures are then converted to equivalent static loads for evaluation or design purposes. Figures are provided to describe the effect of external equivalent static blast loads as a function of explosive quantity, quality and distance in the 1 to 10,000 lbs range and 25 to 2,000 ft distance. Finally, example problems are presented for blast-equivalent TNT phenomena associated with side-on and reflected pressure and their conversion to equivalent static loads for design or evaluation purposes. It should be noted that procedures contained here are simplified approximations. To obtain more accurate response of an SSC it is necessary to consider the time-dependent blast wave transversing the SSC. The result load on the SSC involves not only the characteristics of the blast wave and the global and local response frequency of the SSC, but also the geometry of the structure. With the exception of postulated blast loading applied to petrochemical facilities [1, 2, 3], these types of analyses and design bases are normally applied only to hazardous nuclear facilities as part of the required design or evaluation procedure. In [4] is a detailed review of pipe failure that has occurred in U.S. Commercial Nuclear Reactor Plants between 1961 and 1997. In the 36-year period studied there were 2,247 recorded failure events, which included 118 pipe ruptures reported in the database. As a result of this experience, commercial nuclear power plant facilities are required to consider pipe rupture in high-energy piping and leakage in moderate energy piping as a design basis for the piping and the facility by the U.S. Nuclear Regulatory Commission (NRC). The U.S. NRC has developed definitions for moderate- and high-energy piping [5]. The analysis of fluid distribution systems normally does not consider vibration as a design-basis requirement. However, once the system is put into service vibrations may be observed due to active equipment (pumps or valves) operation or fluid flow, and analysis may be required to determine whether the observed vibration may cause fatigue types of failure of the system involved [1]. Each type of response may correspond to different failure modes; hence, they should be addressed in a specific way corresponding to the type of vibration being observed or considered. The most typical case is represented by the beam-type response as a result of the excitation of the structural mode shapes. Piping system structural vibrations typically correspond to the low-frequency range (less than 20Hz), while ductwork or other thin-walled distribution systems may respond with higher-frequency plate or shell modes of vibration. Buried pipeline systems are commonly used to transport water necessary for heat removal and fire protection safety function for hazardous facilities. In some instances they may also conduct radioactive or other toxic and pathogenic fluids. Safety-related cables such as emergency power supply and instrumentation and control lines between buildings are often routed in buried reinforced-concrete tunnels or vaults. All these types of structures are referred to in this chapter as buried systems. These buried systems are often required to remain functional in case of an earthquake not exceeding a design-basis earthquake. This means that failure modes are considered that may induce large deformation in the pipes, tunnels or vaults; or that may produce pipe, tunnel or vault support failure; or that may produce tunnel or vault wall or roof collapse over the pipes or cables. Hazardous facilities are defined as those facilities that process, store, handle or transport detonable, toxic or pathogenic materials and waste in a form and quantity such that a life-threatening or serious injury hazard potentially exists for collocated workers, the public, or significant damage to the environment. These facilities are divided into three hazard categories as described in Chapter 1. For safety-related structures, systems and components (SSC) in the more hazardous facilities (hazard category 1 and SSC performance categories 4 and 3) characteristic of high-hazard nuclear facilities, a formal written QA program plan should be required. Hazard category 2 and performance category 2 SSC characteristic of petrochemical and biomedical and moderate hazard nuclear facilities also should have a written, somewhat less formal, QA plan. For the lower and PC-1 and PC-0 SSC, a formal written QA plan is not usually required. This is typically referred to as the graded approach to QA. It should be understood the QA usually is the procedures and documentation that ensure that the QC examination and tests contained in applicable industry standards or owner specifications are properly implemented and evaluated. Section 11.2 discusses Safety Classification and a graded and defense-in-depth approach to QA. Section 11.3 discusses quality in the constructed project, Section 11.4 provides observations, conclusions and recommendations, and Section 11.7 provides References. Appendix 11.A provides a detailed comparison of ISO-9001-94 [1] with NQA-1-2000-10CFR50 [2] Appendix B and 10CFR830.122 [3] which define the basic nuclear QA requirements in the U.S. Appendix 11.B of this chapter provides details concerning the existing Nuclear Procurement Issues Committee (NUPIC) and the role this organization plays in auditing nuclear safety-related product, material and services supplied to nuclear power plants in the U.S. Appendix 11.C is a comparison of ISO-9001-94 requirements and ASME NQA-1 requirement prepared by ASME. Appendix 11.D contains a summary of IAEA QA requirements in comparison to British nuclear QA requirements. Appendix 11.E provides a discussion of QA excesses that have provided the basis for less emphasis on document QA and more emphasis on project and facility performance responsibility for QA. Finally, Appendix 11.F is the Forward to a typical QA program that identifies the different levels of QA that are part of a graded approach to QA. It also distinguishes between the format and content references for a QA program manual applicable to a professional services organization which typically follow the 10CFR832.122 [3] or IAEA QA [4] and that which is meant to guide a vendor of nuclear safety related products and services [2] which generally follows the QA format of the USNRC[4]. By far the most formal application of QA has been its application in the nuclear industry. This back matter includes Index

This comprehensive and authoritative volume to decommissioning of nuclear facilities will serve as a complete introduction to those new to the field, as well as an up-to-date desk reference on regulations and resources for experienced practitioners. The handbook will provide a forum building a network of consistent approaches, practices, and results. Covering both NRC and DOE approaches, this book applies not only to decommissioning existing facilities, but by crossing the traditional lines between operations and reuse, this will also allow us to rethink the construction of new ones. The expert team of authors provides valuable lessons from their collective experiences in nuclear decommissioning. They represent areas pertaining to policy, engineering, and science. The handbook focuses primarily on time-tested and proven technologies. The text is supplemented by a CC-ROM, which provide readers with a living resource through links to Internet-based updates containing latest information. This front matter contains the table of contents Decommissioning brings to closure, or terminates, the mission or useful life of a radiological or nuclear facility. The owner or licensee normally decides when a facility is to permanently cease operations. It must then show that a proposed decommissioning project plan can be conducted safely, and that at completion the facility will comply with regulatory requirements. This chapter presents a road map of the decommissioning process, highlighting the major issues to address in planning, implementing, and completing a project, and including references for further detail. During the more than half-century history of the U.S. weapons programs, a vast research, production, and testing network was developed that came to be known as the “nuclear weapons complex..” The DOE and its predecessor agencies have been responsible for most of these facilities. DOE-owned plants and equipment include: - Facilities for uranium milling and refining - Gaseous diffusion plants for producing weapons-grade uranium - Facilities for fabricating and testing weapons - Nuclear reactors for testing materials and equipment components - Nuclear reactor prototypes - Cyclotrons - Hot Cells - Research laboratories They also included reactors for making tritium, plutonium, and radioactive and stable isotopes of many elements; and chemical separation facilities for extracting uranium, plutonium, and other radionuclides. This chapter focuses on decontamination activities at nuclear and radiological facilities regulated by the NRC. If a state is an NRC Agreement State, certain nuclear-related facilities (excluding federal facilities and nuclear power reactors) may be regulated through the relevant individual state agency. The NRC normally accomplishes its regulatory function by issuing and then enforcing the provisions of a license. It licenses and regulates nuclear reactors (both power and nonpower generating) and over 6,000 radioactive material licensees (NRC 2003a). The NRC also regulates several DOE facilities, either by mutual agreement or by specific statute. A relevant example of statute regulation is the licensing of the permanent geologic repository at Yucca Mountain for the disposal of spent nuclear reactor fuel and high-level waste (HLW). This chapter summarizes the major environmental laws relevant to decommissioning. Whereas the Atomic Energy Act (AEA 1954) is the primary law governing the management of radionuclides, the Resource Conservation and Recovery Act (RCRA) is the primary law governing hazardous waste. Other statutes also address the release and disposal of radioactive and hazardous materials. Although the DOE, the NRC, and the Environmental Protection Agency (EPA) are the primary agencies implementing these requirements, much of the authority to develop environmental regulations and enforce them has been delegated to the states. It is not practical to address the individual state environmental requirements in this handbook, many of which are similar to the federal requirements. In general, environmental laws, regulations, and other requirements for decommissioning of facilities under DOE management are interpreted and communicated through a directives system and are imposed on specific projects through contracts. The internal focus is on the final phases of the life cycle, including deactivation, decommissioning, surveillance and maintenance, and any potential transfer of ownership. The management system includes mandatory requirements, such as policy notices, orders, notices, and manuals, and guidance with some implementation flexibility, including the graded application of certain technical standards and guides. In addition, the management system is supported by optional handbooks, voluntary standards, and other documents illustrating established practices for meeting requirements. A thorough understanding of this system requires acceptance that the DOE is a mission-oriented federal agency with its share of statutory intricacies, diverse sites, and a diverse set of contractors. Decommissioning of nuclear facilities requires attention not only to radiation safety but also to other safety requirements collectively referred to as occupational safety and health (OSH). As a general rule, a safety assessment is the initial phase of both radiation safety and OSH. The safety assessment identifies significant hazards during the decommissioning phase of a nuclear facility that are not normally encountered during the operational phase. Hazardous materials are major factors in the decommissioning of old nuclear facilities and represent a risk to the operators undertaking the work. Examples of common hazardous materials are lead, asbestos, PCBs, mercury, and beryllium. All require special disposal. The handling and disposal of mixed (hazardous and radioactive) wastes can pose special problems since, unlike radioactive materials with a relatively short half life, hazardous chemicals may pose a health hazard for a much longer (if not indefinite) time. If decommissioning will be long deferred, due regard should be given to the gradual deterioration of structures, systems, and components. Upon permanent shutdown, a facility should undertake a critical review of all site records (paper, microfilm, and electronic formats) and should resist the initial urge to eliminate records used “for operations only.” Records are critical assets for decommissioning planning; their maintenance and accessibility should preclude any decision as to reductions in related site staffing. Decommissioning normally follows deactivation, and often is followed by long-term storage with S&M. A long time may pass between facility operation, deactivation, and final decommissioning. The S&M activities focus on monitoring and controlling any remaining hazardous substances or contamination and maintaining the facility's structural integrity. In some cases, facility operations may be temporarily suspended, then indefinitely shut down, establishing an S&M phase by default. Decommissioning a nuclear facility requires a multitude of decisions that may have wide ranging effects. One critical aspect to consider in making decisions is the risk of potential adverse impacts on the facility workforce, the public, and the ecosystem. According to the American Association of Engineering Societies, risk analysis may be considered to include three essential modules: assessment, management, and communication. These modules are interrelated, as shown in Figure 7.1 (AAES 1998). The end of a nuclear facility's mission and its transition into decommissioning can create substantial hardship on nearby communities, particularly in the case of major facilities built at a distance from large metropolitan areas. A rural setting often promotes the establishment or growth of smaller communities around the nuclear facility that come to rely heavily on the facility for their economic well-being, either directly through taxes, or more indirectly, by services and support functions to the facility and its employees. Eliminating permanent staff during plant closing can result in significant negative economic effects, although a temporary local economic upturn is possible as specialized contractors, managers, and laborers move to the area for lengthy decommissioning activities. Because of these financial impacts, local community opinion could be anywhere from neutral to negative initially, even if shutdown were announced well in advance. The reaction can be more dramatic in the event of swift or unannounced plant closures. This chapter gives a general overview of public and stakeholder participation. In the past, many agencies and organizations responsible for decommissioning made no distinction among various segments of the public. In effect, they identified stakeholders as anyone wishing to be involved. That approach fails to distinguish the general public from people clearly and unambiguously impacted by decommissioning decisions. Furthermore, such an approach makes it virtually impossible to reach the stakeholders of primary concern. The process described here is based on new developments in stakeholder participation and was used by the Peer Review Committee of the American Society of Mechanical Engineers and the Institute for Regulatory Science (ASME/RSI) and has since been applied in several peer reviews (ASME/RSI 2001 and 2002). Initial decommissioning planning occurs, whether intended or not, before the decision is made to permanently cease operations. The initial planning assessments completed by the facility owner are defined to generate specific and sufficient information, including costs related to continuing operation for specific time periods, ceasing operation by a specific date, and determining the type of decommissioning to pursue and a start date for preliminary activities. Key decisions finalized early in the planning process to make for a safe and cost effective project. There are many examples of nuclear facilities where a determination was made to cease operations for an indeterminate time period, while assessments regarding the facility's future use were completed. Examples include the U.S. Department of Energy's (DOE) Fast Flux Test Facility remaining in hot standby until early 2003 (Lanais 1993), and Millstone Unit 1, a commercial nuclear reactor now in safe storage. In some cases, initial decommissioning studies are performed while a facility is in this shutdown pre-decision mode, A commercial power reactor facility in hot standby requires considerable expenditures to satisfy the U.S. Nuclear Regulatory Commission (NRC) requirements so that the facility might be returned to operation at a reasonable cost without re-testing equipment. This approach is costly since no operational benefits are derived but costs near operating level continue. The costs may be significantly less for small DOE reactors, and may be much less than normal operating costs for non-reactor facilities. Though not required before starting decommissioning activities, determining the ultimate end-state of the facility and the site is a critical decision. Substantially different approaches may be implemented for a facility that will be remediated for potential residential use (greenfield) than for one that is to be partially demolished for potential industrial reuse (brownfield). In the greenfield approach, almost everything with discernible radioactive contamination would likely be removed and shipped off site for disposal. In the brownfield approach there are more options. For example, only some of the high radioactive content structures and components may be dismantled for off site disposal. Only a portion of the above- and below-grade structures might also be removed, along with some of the contaminated soil beneath. While above-grade portions of buildings on a brownfield site may be demolished, their foundations may remain in place with little or no decontamination. Structures intended for reuse might have their internal radioactive components removed and the remaining interior surfaces decontaminated to sufficiently suitable limits to allow unrestricted reuse of the remaining space. This chapter describes the collection of information needed to plan the decommissioning of a nuclear facility. Characterization is frequently thought of as synonymous to radiation measurement, but the term has a broader meaning. Characterization is the collection of all information needed to describe, in adequate detail, the following: - The hazards present at/in the facility - The condition of the facility structure as it may affect worker health and safety - The extent, nature, and concentration of radiological and hazardous chemical contamination - The institutional, legal, and technical restraints on decommissioning alternatives Characterization is an iterative process that precedes and parallels actual decommissioning. The characterization program discussed here is in support of decontaminating structures to protect workers and future site occupants, but this effort is intimately related to the consideration of ground contaminants, beneath and around the structure, and their potential travel to off-site environs. Performing site and facility characterisation assessments is often complex Chapter 10 presented the needs for a facility undergoing decommissioning. As discussed in Chapter 7, a key aspect to risk assessment is the evaluation of radiological exposures to hypothetic My maximum exposed individuals.. Computer-based tools have been developed to help address these challenges. This chapter specifically discusses three computer-based aids developed by government agencies for the conduct of decommissioning, namely RESRAD, DandD, and Compass, together with MARSSIM, and the adaptive characterization approach they support. RESRAD and DandD are used to conservatively calculate exposures to the maximum exposed individual through multiple pathways, based on measured values of residual radioactive contamination, These two codes are the primary dose modeling approaches considered acceptable by the NRC and DOE. Compass is a code used to automate some processes from the MARSSIM characterization protocols. The MARSSIM methodology provides information on planning, conducting, evaluating, and documenting the final status radiological surveys of building surface and surface soils for demonstrating compliance with dose- or risk-based regulations or standards. MARSSIM is a multi-agency consensus document that was developed collaboratively by four federal agencies having authority and control over radioactive materials: the Department of Defense (DOD), the DOE, the EPA, and the NRC. The manual was originally issued in December 1997 as NUREG-1575 (NRC 1997a) and then as NUREG 1575, Rev.1 (NRC 2000a) in August 2000 to incorporate comments received on the 1997 version. Errata and modifications that were included in Revision 1 are shown on page xxviii of that document. In general, the two versions are quite similar; the former is part of the documentation of completed projects and the latter is used for future and on-going projects. This chapter also addresses two approaches to characterization in decommissioning, MARSSIM is widely used to optimize a characterization program while helping ensure statistical validity of the results. The program is focused toward field radiological characterization. The second approach is an adaptive sampling and analysis program that has been demonstrated to improve cost and time performance objectives for characterization (DOE 2001c). The decommissioning cost estimate and schedule do not stand-alone; they are an integral part of the planning for a project from the concept to the final implementation. The cost estimate and schedule are linked inseparably, as changes to the cost affect the schedule, and vice versa. Only with an accurate cost estimate and schedule can management usefully track costs and project trends. Reliable cost estimating is one of the most important elements of decommissioning planning. Alternative technologies may be evaluated and compared based on efficiency and effectiveness, and measured against a baseline cost as to the feasibility and benefits derived. When the plan is complete, those cost considerations ensure that it is economically sound and practical for funding. Estimates of decommissioning costs have been performed and published by many organizations. The results of an estimate may differ because of different work scopes, different labor force costs, different money values because of inflation, different oversight costs, the specific contaminated materials involved, the waste stream and peripheral costs associated with that type of waste, or applicable environmental compliance requirements. Some of the divergence in costs, however, cannot be easily explained, and this lack of consistency prohibits direct estimating by measuring standard quantities, such as initial capital east, facility size (megawatts), square footage of facilities, or volumes of waste streams. At some point, it may be possible to multiply one or more of these measurements by some predetermined number to arrive at a cost estimate. But until such a system is proven to be reliable, a reasonable degree of reliability and accuracy can only be achieved by developing decommissioning cost estimates on a case-by-case basis. The decisions and process for decommissioning nuclear facilities play a key role in controlling radiation exposures. The immediate goals of such projects are to reduce or eliminate occupational and public radiation exposures, primarily by isolating contaminants from surfaces and subsurfaces, and to enable cost effective radioactive waste management and disposal. Decisions that define the applicability of various processes and techniques arc at the heart of decommissioning. Potential techniques include washing, heating, chemical or electrochemical action, mechanical cleaning, and others. For large facilities and equipment, applicable methods are mostly chemical or mechanical. For components that have a role in a facility with a continuing operation mission, dilute chemical agents are typically used to nondestructively dissolve just the outermost surface film bearing most of the contamination. Conversely, if there is no potential for reuse, the base metal can also be dissolved. The process for a small contaminated metal shack may he brief, as compaction may be obviously the most viable option, while a large highly radioactive facility may require more complex processes, including options, such as the wide use of robotics. Intermediate project objectives on the path to restoring the facility to a defined end state include: - Removing loose radioactive contaminants and fixing the remaining contamination - Minimizing residual radioactive contamination requiring protective storage - Maximizing recycle and salvage of equipment and materials - Minimizing volume requiring disposal as radioactive waste - Segregating waste for shipping and disposal in the least restrictive acceptable form (e.g., unrestricted, clean rubble, sanitary, special class, surface contamination only, hazardous, radioactive low-level, mixed hazardous and radioactive, remote handled, and transuranic waste) - Minimizing any period requiring protective storage or long-term monitoring When chemical decontamination is indicated, it may be necessary to temporarily dedicate appropriate facilities for precipitation, filtration, evaporation, demineralization, stabilization, recycling, and or reclamation. Residual concentrates may represent a significant source of radiation before transportation for treatment and/or disposal. Each step is an expense to the budgeted occupational dose and increases the risk of an unplanned release, such as the uptake of radioactive material that results in higher doses than those from handling the contaminated system without extensive decontamination. Planning is therefore the process of selecting options based on a continual and iterative cost-benefit analysis. This chapter presents information related to spent fuel acceptance, management, storage options, and the operational and financial considerations faced before and during decommissioning. The generation of electricity by nuclear fission creates highly radioactive waste in the form of spent fuel. As each reactor is refueled, one-fourth to one-third of the fuel is removed and placed in the plant's storage pool, which shields and cools the spent assemblies. Since these fuel pools were initially designed for provisional storage, most plants cannot store all the assemblies generated over the expected reactor operating life. In the late 1960s and early 1970s, assemblies could be transferred off site for reprocessing, freeing up additional space in the storage pools. It was anticipated that the DOE would provide interim storage at centralized sites. In the meantime, sites took measures to extend their own storage capacity, often by replacing the storage racks in the pools with denser ones to extend capacity as. much as practical. When that was no longer sufficient, supplemental storage was acquired, typically as an independent facility, With additional delays in the development of the government's waste management system, plants shut down over the next ten years will have a full compliment of fuel in on-site storage and will need to address its disposition in planning for decommissioning. The NRC gives no specific guidance on managing or funding such storage. It does, however, refer a licensee to the provisions of 10 CFR 50.54(bb) (NRC 2001b), which require the licensee to submit for NRC approval the program by which it intends to manage and fund the management of all irradiated nuclear fuel at the reactor until possession transfers to the DOE. Whether funding is available as government reimbursements, damage claims, or through rate relief, the owner must determine the magnitude of potential financial liability. Many activities must be performed just before or soon after a nuclear facility concludes its mission and enters the transition from operations to decommissioning. These activities include assurance of immediate and long-term safety for the facility systems, human resources challenges, systems configuration changes, regulatory changes, and stakeholder relations. Long-term stewardship and eventual site closure activities must also be considered and addressed. Characterization is an essential and substantial part of decommissioning. An initial scoping survey must be done to identify personnel hazards and to develop a decommissioning plan. Detailed characterization data must be collected to determine the type and extent of contaminants before any actual decontamination or dismantling. Surveys must be done during decontamination to test the effectiveness of the efforts; they may be either the in-process type or post-decontamination, or both. A final status survey is required, and a confirmatory survey may also be required to document the facility end-state, In addition, should the facility be placed in some form of long-term or interim storage, it will be necessary to characterize the remaining radioactive material and hazardous chemical waste to support long-term stewardship requirements. Such extensive characterization efforts contribute substantially to the overall cost of a decommissioning project. No single technology can address the full spectrum of requirements for facility and material characterization. Decommissioning managers must identify the appropriate suite of characterization methods needed to thoroughly characterize the facility throughout the life of the project. Often, characterization methods are used in combination with one another to take advantage of the strengths and compensate for the limitations of each method. State-of-the-art characterization technologies show one or more of the following competitive discriminators: - Reduced labor to conduct the characterization activities - Improved data management, including automated generation of data reports - Practical application of standard laboratory practices in the field - Improved data accuracy, precision, and detection limits - Deployment of survey devices into inaccessible locations - Reduced radiation/chemical dose by using remotely deployed instruments - Improved operation - Reduced training and education requirements - Real-time or near real-time generation of data - The easily read data, and interpretation thereof - Acceptance of results by regulators Decontamination is the removal of material from areas where it is not wanted. This chapter will deal with the removal of radionuclides, but the same or similar technology could be used for the removal of hazardous chemical material. The technology is similar to the cleaning of dirt, oil. or corrosion products except that radionuclides are associated with the material, Both cleaning and decontamination require similar technologies, methods, equipment, and procedures but the degree of cleaning may be different. The degree of cleaning is defined by certain cleanliness standards. Decontamination is based on cleanliness standards that typically revolve around a personal dose rate or contamination level associated with the component or surface being cleaned. For disposal of the cleaning material, the chemical composition of both the waste and the radionuclide content must be understood. The disposal method and packaging will depend on all of these items. Decontamination is used to reduce the dose that workers may receive from a component or surface, to reduce the potential for airborne radionuclides, or to reduce the associated disposal cost. Some decontamination can allow for the reuse or recycle of the material, although this approach is currently less common. An example of reuse of material is the melting of slightly contaminated steels to be manufactured into shield blocks for a particle cyclotron. The shield blocks keep natural radiation sources from interfering with the acceleration of particles. Many tons of materials have been recycled for use in nonradioactive environments. In order to accomplish a use for “free-release,” the material must not only be cleaned but also surveyed and found to be below the releasable level of the facility. The facilities that perform “free-release” are licensed by the federal government or the state in which they operate. The license and regulations specify the release limits of the facility. This document is meant as an introduction to decontamination, both chemical and mechanical, with references to direct the reader to greater detail. The listing of any commercial companies should not be construed as a recommendation for the offerings of products or services. 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. Decommissioning and remediation projects demand an integrated, life-cycle approach from planning through waste management. Concepts that must be clearly understood in developing a comprehensive waste management program include the various categories of waste, waste minimization, regulatory issues, and differing management approaches. As with most aspects of decommissioning, waste management provides opportunities for optimization. No one solution works in all situations and facilities, but understanding the differing approaches will let the manager make prudent and efficient decisions to best support stakeholder needs. Management of tritium waste is sufficiently unique that it is discussed separately in Section 19.7 Many U.S. nuclear power plants have permanently shut down and are in some phase of decommissioning. They include research reactors, other research facilities, and a small number of other commercial nuclear facilities, such as three commercial spent fuel reprocessing plants. There are hundreds of DOE nuclear facilities, i.e., reactors, fuel reprocessing plants, research facilities, and laboratories that supported a wide variety of operations using hazardous substances in the production of nuclear weapons. The DOE has restored many of its nuclear sites and cleaned up many hazardous waste sites. The decommissioning of commercial, private, and DOE-owned sites has resulted in the development of effective cleanup technologies and practices. Radioactive and hazardous waste constituents found in the soil and groundwater during decommissioning got there because of spills, system upsets, maintenance activities, and outdated work practices. Only a few chemical processes support electricity production at commercial power reactors, but nevertheless, activities at these sites released contaminants to the environment. An example of such an activity is work performed in on-site analytical laboratories. These laboratories were often the source of radioactive and chemical agents that reached the environment via sink and floor drains. At some of the older facilities, both radioactive and chemical contaminants were found in leach fields where the contaminants tended to accumulate. The primary radioactive contaminants in the soil or groundwater at a commercial site are 137 Cs, 90 Sr, 60 Co, and 3 H. The source of tritium in the groundwater is. generally water leakage from the spent fuel storage pool that tends to occur over many years of operation, Insoluble radioactive contaminants tend to be trapped in the soil and are not easily transported by groundwater movement, therefore tend to remain in localized areas. The Maine Yankee nuclear electrical generation plant is located on an 820-acre site in Lincoln County in the Town of Wiscasset, within the mid-coast region of the State of Maine, about 27 miles northeast of Portland. The plant is owned by a consortium of ten New England electric utilities and was operated by the Maine Yankee Atomic Power Company (MYAPCO). During its 24-year operating lifetime, Maine Yankee generated and distributed more than 125 million MW-hours of electrical power. The construction permit for Maine Yankee was issued on October 21, 1968. The operating license, issued on September 15, 1972, authorized operation of the pressurized water reactor (PWR) facility until October 21, 2008, with power levels up to 75% rated thermal power. Commercial operation began on December 28, 1972. The license granted in June 1973 allowed for operation at 100% rated thermal and at power levels up to and including 2440 MWt. Rancho Seco was a single unit, 913-MW electrical, Babcock & Wilcox design, nuclear power plant owned by the Sacramento Municipal Utility District (SMUD). The plant is 25 miles southeast of Sacramento, California on 2,480 acres of ralling terrain. Commercial operation of this PWR began in 1975 and ceased in 1989, The NSSS, housed within a hardened containment building, consisted of two independent primary coolant loops (each containing two reactor coolant pumps and a steam generator), an electrically heated pressurizer, and connecting piping. Major structures included two cooling towers, reactor building, administration building, auxiliary building, training and records building, nuclear service electrical building, interim onsite (waste) storage building, and support buildings (security, warehouses, and an unfinished technical center). An ISFSI was added in 1995. The Trojan Nuclear Plant is located in Columbia County, Oregon, about 42 miles north of Portland. The 634-acre site includes a recreational area/park, various office buildings, and an industrial area enclosed by a security fence. Portland General Electric (PGE) is the majority owner and has responsibility for operating and maintaining Trojan. The Bonneville Power Administration, a power marketing agency under the DOE, is obligated through net billing agreements to pay costs associated with Eugene Water and Electric Board's share of Trojan, including decommissioning and spent fuel management costs. Big Rock Point, constructed on the shores of Lake Michigan near the resort town of Charlevoix, Michigan, began operation as a research and development reactor. In January 1960, Consumers Power (later Consumers Energy) submitted its formal proposal for a construction permit and operating license. The permit was granted four months later. Ground breaking for the plant took place July 20, 1960, and the plant was officially completed by September 1962. Big Rock Point was a relatively small plant (67 MWe) that made large contributions to the nuclear industry. As the world's first high power-density boiling water reactor (BWR), one goal was to demonstrate that nuclear plants could generate electricity economically. In 1977, Big Rock Point set a world record for BWRs by operating for 343 consecutive days. The Haddam Neck Plant is located in Middlesex County on the east bank of the Connecticut River, about 21 miles south of Hartford, in Haddam Neck, Connecticut. The Connecticut Yankee property consists of 563 acres, of which 4.5 acres are fenced-in power plant structures. Most of the property remains open forest. The San Onofre Nuclear Generating Station (SONGS) Unit 1 is located on the coast of Southern California, in San Diego County, next to operating Units 2 and 3, The SONGS site is located entirely within the Camp Pendleton Marine Corps Base under an easement granted by the U.S. Government. Southern California Edison (SCE) and San Diego Gas & Electric (SDG&E) are joint owners. The UTR-10 was an Argonaut-type reactor using uranium enriched to 19,75% in 235U in a graphite-reflected, water-moderated core. It was designed and built in 1959 by the Advanced Technology Laboratories division of the American Radiator and Standard Sanitary Corporation. Initial reactor criticality came on December 31, 1959; operations officially ceased on May 15, 1998. The reactor was housed in the Nuclear Engineering Laboratory located on the west edge of the main campus of Iowa State University (ISU) in Ames, Iowa. The facility is a two-story, three-level, brick building built in 1934 by the Department of Agriculture. It was deeded to the University in 1946. The reactor room housed the reactor (enclosed in a concrete biological shield), the process pit, the fuel storage pit, and a five-ton bridge crane. The Barnwell Nuclear Fuel Plant (BNFP), located in Barnwell, South Carolina, on a 1622 acre property about five miles southeast of the Savannah River site, was built in the early 1970s to reprocess spent nuclear fuel from commercial power reactors but was never used for this purpose. However, extensive testing used natural uranium as a surrogate material, and research and development work in the laboratories used plutonium and other transuranics (TRU). In 1942, the U.S. Army Corps of Engineers chose Hanford as the location for reactor, chemical separation, and related facilities for producing and purifying plutonium. A small farming and ranching community along the Columbia river, covering the towns of Richland, Hanford, and White Bluffs, was designated as site W of the Manhattan Project, the code name for the government's development of atomic and the nuclear weapons.. The land known as the Hanford Engineering Works was located in a sparsely populated area close to the Bonneville and Grand Coulee dams. This back matter contains index

A completely revised and updated edition of the classic and comprehensive guide to the construction rules for power boilers—their intent, application, and interpretation. This unique guide provides expert advice and useful information for design engineers, project managers, architect engineers, manufacturing engineers, boiler operators, insurance inspectors, and other power boiler professionals. This edition includes explanation and use of the other Sections of the ASME Boiler and Pressure Vessel Code that affect construction. It contains chapters on boiler life extension and repairs and alteration of boilers under the rules of the National Board Inspection Code. Current to the 2010 Edition of Section I. Founding Authors: Martin D. Bernstein and Lloyd W. Yoder. This front matter contains the table of contents, list of figures, list of tables, foreword, introduction and acknowledgement. Click on PDF icon to view complete front matter. Section I applies to several types of boilers and components of boilers, such as economizers, superheaters, reheaters, and, in some circumstances, feedwater heaters. Although its title is Rules for Construction of Power Boilers, its scope is somewhat broader. The Preamble to Section I explains that it covers electric boilers, miniature boilers, high-temperature water boilers, heat recovery steam generators, certain fired pressure vessels, and organic fluid vaporizers. Since the precise definitions of these various types of boilers are not generally known, the following definitions, found in footnotes to the Preamble, are helpful: Power boiler —a boiler in which steam or other vapor is generated at a pressure of more than 15 psi (100 kPa) for use external to itself. Electric boiler —a power boiler or a high-temperature water boiler in which the source of heat is electricity. Miniature boiler —a power boiler or high-temperature water boiler in which the limits in PMB-2 are not exceeded. (PMB-2 establishes limits on shell diameter, heating surface, gross volume, and maximum allowable working pressure. See further discussion under Part PMB later in this chapter.) High-temperature water boiler —a water boiler intended for operation at either pressures in excess of 160 psi (1.1 MPa) or temperatures in excess of 250° F (120°C), or both. Heat recovery steam generator (HRSG) —a boiler that has as its principal source of thermal energy a hot gas stream having high-ramp rates and temperatures such as the exhaust of a gas turbine. Fired pressure vessel —reheater, isolable superheaters, and nonintegral separately fired superheaters. Section I is a set of rules for the construction of boilers. These rules are not mandatory unless invoked by contract or adopted as law by a governmental entity, usually referred to as a jurisdiction. A jurisdiction can be a state, province, county, or city. If there happen to be no jurisdictional requirements regarding Code coverage, the purchaser of a boiler may specify to the manufacturer construction to Section I rules as a requirement. Currently, Section I has been adopted by all but one state (Wyoming is the exception) and by all of the provinces of Canada. Wyoming has a pressure vessel law, but has not enacted a boiler law nor developed boiler rules and regulations. Because of Wyoming's relatively sparse population and limited industrial facilities, the legislature apparently does not consider the expense of implementing the law to be justified, since as a practical matter, all boilers in the state probably meet the ASME Code. Seventeen cities, counties and territories also have laws requiring Section I construction. States and provinces with boiler laws typically enforce the Code by means of an appointed board or commission that writes and revises boiler rules and an appointed chief inspector with the necessary enforcement staff. The jurisdictional authorities also depend on the efforts of Authorized Inspectors working for Authorized Inspection Agencies, whose function is to assure that the boiler manufacturer has complied with Section I. These same jurisdictional authorities also have the responsibility for overseeing the care and operation of boilers once they are completed. Material is a fundamental link in the chain of ASME Code construction, and great care is taken to assure its quality. The ASME accomplishes this by adopting material specifications that have been developed and adopted first by the American Society for Testing and Materials, the ASTM. The ASME material specifications are thus usually identical to those of the ASTM. ASTM issues specifications designated by letter and number, for example, A-106, Seamless Carbon Steel Pipe for High Temperature Service. When the ASME adopts ASTM specifications, it adds the letter S. The equivalent ASME pipe specification is thus SA-106. In keeping with the ASME's policy to internationalize its Code, ASME has adopted a number of foreign material specifications. These are included in Section II, Part A as a single cover sheet for each specification (a complete material specification has not been reproduced in Section II, Part A, as is done for the ASTM material specifications). For example, Section I permits the use of SA/AS 1548, Steel Plates for Pressure Vessel Equipment, an Australian Standard Material Specification. If a boiler manufacturer chooses to use one of the permitted foreign material specifications in the construction of his Section I component, then he will be obliged to obtain a copy of the relevant material specification. The ASME or SA material specifications are used as purchase specifications. Each specification contains a variety of information appropriate to that product, dealing with how it is ordered, the manufacturing process, heat treatment, surface condition, chemical composition requirements, tensile requirements, hardness requirements, various test requirements, and how the material is to be marked. The purchaser can also specify nonmandatory supplementary requirements, dealing with such things as stress relieving, nondestructive examination, and additional testing. All of these requirements have evolved over the years in response to the needs of the users. The Foreword to Section I states that the objective of the rules is to afford reasonable protection of life and property and to provide a margin for deterioration in service so as to give a reasonably long safe period of usefulness. Just how long is that period? There are many 40-year-old boilers that are still giving good service, and sometimes, 40 years is cited as the expected life of a boiler, assuming of course that it is properly maintained. However, Section I gives no explicit indication of what constitutes a reasonably long design life. Sometimes, the criteria used to establish allowable design stresses at elevated temperature are mistakenly assumed to imply a limited design life of only 100,000 hr (about 14 yr if the boiler is in service 80% of the time). These criteria are found in paragraph 1–100(b) of Appendix 1 of Section II, Part D, from which the following excerpt is taken: At temperatures in the range where creep and stress rupture strength govern the selection of stresses, the maximum allowable stress value for all materials is established by the Committee not to exceed the lowest of the following: (1) 100% of the average stress to produce a creep rate of 0.01 %/1000 hr. (2) 67% of the average stress to cause rupture at the end of 100,000 hr. (3) 80% of the minimum stress to cause rupture at the end of 100,000 hr. The criterion that usually governs is the second, 67% of the average stress to cause creep rupture at the end of 100,000 hr of service. Note that this stress is not one that would cause creep rupture after 100,000 hr of service; it is only 67% of that stress. Creep life (i.e., the time it takes to cause creep rupture in elevated temperature service) is a complex nonlinear function of stress and temperature. By experiment and analysis, the relationship among temperature, stress, and time to creep rupture has been established and plotted as a series of creep rupture curves. These are log stress versus log time curves, with temperature as a parameter. When plotted this way, the “curves” are almost straight lines. As explained in Chapter 4, under the heading Distinction Between Boiler Proper Piping and Boiler External Piping, Section I rules for the design of boiler proper piping are somewhat limited. They actually cover the design of pipe for internal or external pressure only. More extensive rules are provided in B31.1, Power Piping Code, applicable to the design of boiler external piping. Those rules cover, in addition to pressure, many other loads that piping might be subject to, such as mechanical loads that may be developed due to thermal expansion and contraction of the piping, impact loading, gravity loads, and seismic loads. Although such comprehensive rules are not provided by Section I for the design of the boiler proper piping, the boiler manufacturers over the years have managed to design this piping so that it serves its purpose in a satisfactory manner, often using design methods of B31.1. Designers of boiler piping, whether boiler proper piping or boiler external piping, are thus faced with a mixed bag of interrelated rules and should be aware of a number of special provisions and potential pitfalls that need to be considered. These are discussed in the following sections of this chapter. Boiler external piping and boiler connections are described in PG-58. There are about 20 paragraphs explaining and defining BEP systems for drum-type boilers, isolable and separately fired superheaters, multiple boilers connected to a common steam header, high-temperature water heaters and forced flow steam generators. Section I provides five figures showing the schematics of the Code jurisdictional limits of piping for various piping systems. Fabrication is not specifically defined in Section I, but it is generally understood to mean all those activities by which the manufacturer converts material (plate, tube, pipe, etc.) into completed boiler components. Such activities include welding, bending, forming, rolling, cutting, machining, punching, drilling, broaching, reaming, and others. Section I permits the manufacturer very broad latitude in fabrication because of the wide variations in manufacturing practice, machinery, and trade methods. Section I provides only a limited number of general rules for fabrication in PG-75 through PG-82. (One widely used method of fabrication, tube expanding, has surprisingly few rules provided in Parts PWT and PFT. Welding, perhaps the most widely used and controlled method, is covered extensively in Part PW. A discussion of both methods is given in the next section of this chapter.) The general rules for fabrication in Part PG specify tolerances for drums and ellipsoidal heads and limit certain cutting, drilling, and punching processes such that subsequent machining will remove all metal whose mechanical and metallurgical properties have been affected by the earlier process. The manufacturer is permitted to repair defects in material, provided acceptance of the Authorized Inspector is first obtained, both as to method and extent of repairs. There are several related terms applied to Code construction that are somewhat imprecisely used. These are examination, inspection, and testing. Within the context of Section I, examination usually describes activities of the manufacturer; inspection refers to what the Authorized Inspector does; and testing refers to a variety of activities, usually of the manufacturer. Some confusion arises from the fact that nondestructive examination (NDE) includes activities defined as examinations, but which are often called tests. NDE is an indispensable means of assuring sound construction, since when properly used, these examinations are capable of discovering hidden flaws in material or welds. The examinations referenced by Section I in the NDE category are radiographic examination, ultrasonic examination, magnetic particle examination, and liquid penetrant examination (also called dye penetrant examination). Although not explicitly mentioned in Section I, the use of visual inspection is certainly implied and is called out in B31.1 for application to the boiler external piping. These examinations are often referred to in the industry by the shorthand terms RT (radiographic test), UT (ultrasonic test), MT (magnetic particle test), PT (penetrant test), and VT (visual inspection). Section I generally follows the Code practice of referring to Section V, Nondestructive Examination, for the rules on how to conduct the various examinations, but provides its own acceptance standards. For example, PW-11 calls for certain welds to be examined in accordance with Article 2 of Section V, to the acceptance standard of PW-51. Similarly, PG-25, Quality Factors for Steel Castings, calls for magnetic particle or dye penetrant examination of all surfaces of castings in accordance with Section V, but provides acceptance criteria right there in PG-25. What is third party inspection and why is it mandatory for Section I construction? In the simplest terms, third party inspection refers to the system evolved by the Code committee to assure that code rules are actually followed. That is a basic truth, but it does not fully answer the question regarding why it is mandatory. It is not mandatory for other code constructions such as that governed by ASME B31.1, so why is it so for Section I construction? The answer is twofold; (1) to assure consistent, uniform application of code rules thus assuring a certain degree of safe equipment as well as easing interstate and international trade and (2) to protect the use of the ASME marks (e.g., Code symbol stamps). In the early days of the ASME boiler code, it was not uncommon for each state or jurisdiction to have its own boiler inspection requirements. Because there were no uniform requirements between the states for stamping of boilers, or uniform system for qualifying inspectors and applying Code rules, a boiler manufacturer in one state often had to have its boiler inspected and accepted by the inspector from the state in which the boiler was to be installed. This was a big burden to commerce in general and the boiler manufacturers in particular. It could be argued that this also impacted safety of the equipment due to the differing degrees of expertise of the inspectors regarding understanding and application of Code rules. These conditions led to the formation of the National Board of Boiler and Pressure Vessel Inspectors (National Board) which, with the support of the American Boiler Manufacturers Association (ABMA) and the Uniform Boiler and Pressure Vessels Laws Society, brought about a system for uniform examination of inspectors, consistent application of Code rules and uniform rules for stamping of boilers (i.e., the use of the ASME Code symbol, or mark). This greatly eased the manufacturers' burden of obtaining multiple inspections and stampings and thus making interstate commerce much more efficient. It also added to the general welfare by assuring the public of the uniform application of Code rules. In order for the Code to be effective, there must be some means of assuring that it has been followed. Each phase of the work of designing, manufacturing, and assembling the boiler or component to which the ASME Code symbol is applied must be done in accordance with the Code (see Chapter 8, Third Party Inspection, regarding CSP-53, Protection of The ASME Mark). Assurance that a boiler is constructed to the Code is provided, in part, by a process called Code certification. In the simplest case of a complete boiler unit made by a single manufacturer, the manufacturer must certify on a form called a Manufacturer's Data Report Form that all work done by the manufacturer or others responsible to the manufacturer complies with all requirements of the Code. When some portion is performed by others not responsible to the manufacturer, the manufacturer must obtain the other organization's Code certification. In addition, the manufacturer must stamp the boiler with a Code symbol, which signifies that it has been constructed in accordance with the Code. Certification thus is an integral part of a quality assurance program which establishes that: 1. The organization that did the work held an appropriate ASME Certificate of Authorization to use a Code symbol; 2. The organization has certified compliance with the Code rules by signing and furnishing the appropriate Manufacturer's Data Report Form; 3. The organization has applied the Code symbol stamp to identify the work covered by its Data Report Form; 4. An Authorized Inspector or Certified Individual has confirmed by signing the Data Report Form that, to the best of his or her knowledge, the work complied with the applicable Code rules. The 1960s was a time of rapid growth for the nuclear power industry. In 1963, a new ASME Code section, Section III, was published to cover construction of nuclear power plant components. One of the features of construction under Section III was a requirement that the manufacturer implement a quality assurance system, a formal system of oversight and record keeping assuring that construction takes place as intended, that the right material, parts, drawings, and procedures are used. The benefits of such a system soon became apparent. In 1973, the ASME Boiler and Pressure Vessel Committee (now the BPV Standards Committees) adopted similar but less comprehensive and detailed rules for what is called a quality control (QC) system for the boiler and pressure vessel sections of the Code, Sections I, IV, and VIII (and now Sections X and XII). In Section I, these rules are found in PG-105.4 and Appendix A-300, both entitled Quality Control System. Each manufacturer is required to have a documented quality control system that is fully implemented into its manufacturing operations. It may be of interest to note that the B31.1 Power Piping Code has no provisions for a quality control system that applies to other than boiler external piping, although the governing committee has been considering the idea for many years. The QC system is intended to control the entire manufacturing process from design to final testing and certification. The scope of such systems may vary significantly among manufacturers, since the complexity of the work determines the program required; however, the essential features of the QC system are the same for everyone. Each manufacturer is free to determine the length and complexity of its Quality Control Manual, making it as general or detailed as desired. It is not advisable to include detailed practices in the manual that are not actually followed in the shop. Also, when shop practices are revised, the manual must be updated to reflect the revisions. Otherwise, the Authorized Inspector is likely to find nonconformities, which will cause considerable extra work before full compliance with the Code can be demonstrated. Implicit in the Code QC requirements is recognition that each manufacturer has its own established practices, and these are acceptable if a documented system can be developed showing control of the operation in conformance with Code requirements. It is also recognized that some of the information included in a written description of a quality control system is proprietary, and no distribution is required other than to the Authorized Inspector. The rules dealing with standard pressure parts and how they may be provided are found in PG-11, Miscellaneous Pressure Parts. Ordinarily, Section I requires all pressure parts to meet applicable requirements for material, design, fabrication, welding, inspection, testing, and certification by stamping and data reports. These comprehensive Code construction requirements are considered necessary to ensure safe construction. Considerable relief from this burden is granted in the case of what are called standard pressure parts, provided these parts are designed and manufactured in such a way as to provide an equivalent assurance of quality and safety. Under these circumstances, mill test reports, inspection by an Authorized Inspector, and Partial Data Reports are not required. Among the parts included in the category of standard pressure parts are such components as: pipe fittings, valves, flanges, welding necks, welding caps, manhole frames and covers, and the casings of certain pumps. Most of these parts are readily available as stock items in a variety of sizes. They do not have to be designed for each specific application because the manufacturer has already established pressure and temperature ratings for them. Section I establishes several categories of these parts. One category covers pressure parts which are cast, forged, rolled, or die-formed (i.e., not welded). Within this category, Section I designates two types: parts that comply with an ASME (formerly ANSI and later ASME/ANSI) Standard accepted by reference in PG-42 and parts that comply with a parts manufacturer's standard. There are some differences between the two in material and design requirements. Parts complying with a parts manufacturer's standard must be made of material explicitly permitted by Section I, but parts complying with an ASME standard may, in addition, be made of materials listed in that standard, provided Section I does not prohibit or otherwise limit the use of such materials. The ASME standard also establishes the basis for their pressure and temperature rating and specifies the marking on them; these ratings and markings are accepted by Section I. Some difficulties have arisen in recent years if the standard pressure part is not made of a material explicitly approved for Section I construction in Section II, Part D. This situation is explained in Chapter 3, Materials. The ASME Boiler and Pressure Vessel Code was developed between 1911 and 1914 as a set of safety rules to address the serious problem of boiler explosions in the United States, which at that time were almost a daily occurrence. Average steam pressure in boilers in those days had reached about 300 psi (2 MPa), and the explosions took a heavy toll of lives and property. The principal design basis for boilers and pressure vessels is the safe containment of design pressure. Protection against overpressure was therefore a very important aspect of pressure vessel design. Accordingly, that first edition of what is now Rules for Construction of Power Boilers, Section I of the Code, included rules for the overpressure protection of boilers, based on the best industry practice at the time. The principles of today's Code rules for overpressure protection are little changed from those of the first Code. That first ASME Code of 1914 has now grown into 12 book sections, covering a wide variety of subjects. Six of these book sections deal specifically with the design of pressure vessels and include requirements for overpressure protection. The various pressure-relief devices used for overpressure protection may be categorized as pressure-relief valves (commonly called safety valves), and nonreclosing devices, such as rupture disks. The function of a pressure-relief device is to open at a specified pressure and pass a sufficient amount of fluid to prevent pressure in the vessel or system being protected from exceeding an allowable overpressure above the design pressure. In some book sections such as Sections I and VIII, the design pressure is referred to as the maximum allowable working pressure (MAWP) . Each book section provides rules for determining the required relieving capacity of pressure-relief devices, for establishing allowable overpressure, for determining the maximum set pressure of the pressure-relief devices, and for ensuring that pressure-relief devices will function as required should an overpressure condition occur. Although fundamentally similar, the rules vary somewhat in detail and complexity, reflecting the nature of the application and the somewhat different design philosophies of the committees governing the various book sections. It happens that the Code offers only limited explanation of its rules on overpressure protection, and aside from certain functional requirements, very few rules to guide the designers of pressure-relieving devices. The objective of this chapter is to summarize the Section I rules and to provide some explanation and discussion of them and how they are applied. Since similar rules apply to all the non-nuclear book sections, a brief explanation of those other rules is included, to point out some differences. Thus, the discussion which follows covers all the non-nuclear book sections, with the major emphasis on the rules in Section I. The designer of a boiler must provide adequate cooling flow of water or steam through the various pressure parts so that they will not exceed their intended design temperatures. Water-cooled furnaces must contain high-temperature combustion gases and cool them sufficiently to pass through closely spaced rows of heating surface without overheating that surface or (for coal-fired boilers) causing slagging problems in the furnace or elsewhere. As discussed in Chapter 3 under Special Concerns, Ferritic alloys (carbon and low alloy steels) are used for water-cooled furnace tubes because of their resistance to chloride-induced stress corrosion cracking. Boiler circulation must be sufficient to cool the furnace walls to temperatures at which ferritic steels have adequate strength and oxidation resistance. Section I has never provided rules for boiler circulation, as the responsibility for functional design has traditionally remained with the boiler manufacturer. However, Section I does have rules that implicitly recognize the serious consequences resulting from the loss of adequate water supply to heat-absorbing circuits critical to the safe operation of the boiler. Firing an inadequately cooled or partially dry boiler can lead to overheating of the pressure parts, with consequent loss of strength and potentially catastrophic failures due to internal pressure. Section I addresses this problem with rules intended to assure the availability of feedwater and the presence of water in the drum. The Foreword to Section I states that the objective of the Section I rules is “to afford reasonable protection of life and property and to provide a margin for deterioration in service so as to give a reasonably long safe period of usefulness.” Sooner or later, almost all equipment wears out, and boilers are no exception. The factors that cause the gradual deterioration of a boiler are many. Among them are erosion and corrosion, which cause a gradual wastage and thinning of pressure parts. Two other significant factors that can limit the life of a boiler are creep and fatigue, processes that can cause progressive deterioration of material properties without wastage and thinning. This deterioration is characterized as creep damage or fatigue damage. For practical purposes when discussing boilers, it can be said that creep damage occurs only at elevated temperature, about 700°F (370°C) and above for carbon and low alloy steels and 800°F (425°C) and above for the high alloy steels typically used in superheater construction. Fatigue damage can occur over the entire range of boiler operating temperature, but as will be explained later, components in elevated-temperature service are more susceptible to fatigue damage. These two damage mechanisms are discussed separately, but their effects are additive. Each section of the Code uses design criteria and design stresses appropriate for the components that it covers. The basis for establishing allowable stresses is related to the design philosophy, the type of construction permitted, the degree of analysis required, and the amount of nondestructive examination mandated. Sections I, IV, and VIII, Division 1, use a design approach sometimes called design-by-rule. Design-by-rule is not based on detailed stress analysis. Instead, the rules generally involve the calculation of average membrane stress across the thickness of the walls of the vessels. That is not to say that these sections of the Code completely ignore other types of stress. There is a classic statement found in UG-23 of Section VIII, Division 1, which provides a rationale for design-by-rule using membrane stress alone: “It is recognized that high localized discontinuity stresses may exist in vessels designed and fabricated in accordance with these rules. Insofar as practical, design rules for details have been written to limit such stresses to a safe level, consistent with experience.” In the 1970s, the first sentence was slightly different. At that time the term “high localized and secondary bending stresses” was used instead of the current description “high localized discontinuity stresses.” In either case, bending stress and the effects of stress concentrations are not calculated (at least not explicitly) in Section I. This means, in effect, that the Code assigns allowable membrane stresses with sufficient margin (safety factor) to cover high localized stresses in normal circumstances. As explained in Chapter 9, the signing of the completed Manufacturer's Data Report by the Authorized Inspector and the stamping of the boiler with a Code symbol stamp mark the completion of the boiler. The manufacturer has met its responsibility to supply a boiler which complies with the rules of Section I. Since Section I applies to new construction only, by its own definition in the Foreword, its direct applicability to the boiler has ended. It then becomes the responsibility of the user to maintain the boiler in accordance with requirements of the jurisdiction where the boiler is installed. As can be seen on the map in Chapter 2 showing jurisdictions in the United States and Canada with boiler and pressure vessel laws, all but one state (Wyoming) have boiler laws. For boilers in service, requirements for inspection, maintenance, repair, and alteration vary considerably. Some jurisdictions have specific requirements or invoke those of the National Board Inspection Code (NBIC), which is described below. Unfortunately, however, there are instances each year when lack of proper operating, maintenance, or repair procedures leads to serious accidents. The occurrence of such accidents in the 19th century prompted early legislation to promote boiler safety and eventually culminated in the development of the first edition of Section I in 1915, which comprised what were, in effect, safety rules for the construction of steam boilers. As early as 1833, Congress passed a law establishing a steamboat inspection service requiring steamboat boilers to be inspected every 6 months. Since those early days of steam, the inspector has played a key role in promoting and assuring boiler safety. Gradually, the various jurisdictions that passed laws regulating boilers established organizations to administer and enforce those laws, which typically cover both new construction and boilers in service. Recall now the advice in the Foreword to all sections of the ASME Code that its design rules provide a margin for deterioration in service so as to give a reasonably long, safe period of usefulness. Accordingly, it is the general industry practice that the Authorized Inspection Agencies, insurance companies and jurisdictions continue to accept and insure boilers in service where pressure part wall thicknesses have been reduced below their original design values. In such cases, the pressure parts are monitored to establish the rate at which the walls are thinning, and repair or replacement is accomplished before the pressure part deteriorates beyond an acceptable point. It is difficult to find published recommendations regarding what minimum wall thickness is still safe (likely because of the potential liability involved), so the decision on when to replace a deteriorated component is one normally made by the boiler owner in conjunction with his Authorized Inspection Agency or insurer and the jurisdiction. This back matter contains the references, appendices, index of interpretations, and index.

“Dr. Pioro['s] reputation is excellent. Dr. Duffey has been a leader in the nuclear industry for about three decades…” —Jacopo Buongiorno, Massachusetts Institute of Technology (MIT). “The book would be useful to engineers and scientists working with supercritical fluid flow…there is no comparable text on the market.” —Vijay Chatoorgoon, Department of Mechanical Engineering, University of Manitoba. “I was impressed with the wealth of data and references cited by the authors.” —Maan H. Jawad, President, Global Engineering & Technology. This front matter contains the table of contents, preface, acknowledgments, symbols and abbreviations, glossary and definitions of selected terms and expressions related to critical and supercritical regions specifics & about the authors. The objective of this monograph is to show what work has been done worldwide in the area of heat transfer and hydraulic resistance of fluids at critical and supercritical pressures for the last five decades, including achievements in development concepts of SuperCritical Water-cooled nuclear Reactors (SC-WRs) and in operation of supercritical power-plant “steam” generators. This monograph is the most exhaustive literature survey on this topic published so far. However, some reviews found in the open literature may complement materials presented in this monograph providing in-depth analysis on some specific topics. Therefore, short descriptions of these reviews are listed in Appendix A. This monograph will help the reader to understand the main directions of investigations conducted so far, the problems encountered during the operation of supercritical units and development concepts of SCWRs, and will provide with practical recommendations for the calculation of heat transfer and hydraulic resistance at critical and supercritical pressures. Heat transfer at supercritical pressures is influenced by the significant changes in thermophysical properties at these conditions. For many working fluids, which are used at critical and supercritical conditions, their physical and thermophysical properties are well established (Kirillov 2003). This is especially important for the creation of generalized correlations in non-dimensional form, which allows the experimental data for several working fluids to be combined into one set, as well as the use of numerical solutions (Polyakov 1991). The most significant thermophysical property variations occur near the critical and pseudocritical points (Supercritical Fluids 2002). In general, at the critical point (Kaye and Laby 1973) ( ∂ p ∂ V ) T = T C T = 0 and ( ∂ 2 p ∂ 2 V ) T = T C T = 0 . At temperatures above the critical temperature, a fluid cannot be liquefied. The supercritical fluid is neither a gas or a liquid, and approximates the behavior of a perfect gas. To account for the actual intermolecular effects at high densities, it is necessary to correct the equation of state and physical properties for non-ideal departure from perfect gas values. The equation of state of a fluid can be accurately expressed by a semi-empirical compressibility factor equation of the form: z = p V m R T = 1 + B V m + C V m 2 + … where R is the molar gas constant and B, C , … are the second and third, virial coefficients, which are functions of temperature only, and higher order correction terms also exist. Early studies in Russia related to the heat transfer at supercritical pressures started in the late 1940s. In the 1950s, Podol'sk Machine-Building plant “ЗиО” (“ZiO”) (Plant by the name of S. Ordzhonikidze) manufactured several small experimental supercritical “steam” generators for research institutions such as: (i) “BTИ” (“VTI”)—All-Union Heat Engineering Institute (Moscow) with “steam” parameters of 29.4 MPa and 600°C (Shvarts et al., 1963); (ii) “ЏКTИ” (“TsKTI”)—Central Boiler-Turbine Institute by Polzunov (St.-Petersburg); and (iii) Kiev Polytechnic Institute with “steam” parameters of 39 MPa and 700°C (Kirillov, 2001). The implementation of supercritical power-plant “steam” generators in Russia (the former USSR) started with units having thermal powers of 300 MW (Ornatskiy et al. 1980). Two leading Russian manufacturing plants: “TКЗ” (“TKZ”)—Taganrog Power-Plant Steam Generator's Manufacturing plant (Taganrog, Ukraine) and “ЗиО” (Podol'sk, Russia) developed and manufactured the first units, with the assistance of research institutions such as “ЏКTИ” and “BTИ.” Supercritical “steam” generators are usually the once-through type boilers (Belyakov 1995). Power-plant “steam” generator TΠΠ-110 (TPP-110) manufactured at “TКЗ” in 1961 was the first industrial unit operating at supercritical conditions in the former USSR, and was used at coal-fired power plant. Its design included a liquid slug drain. A total of six units were put into operation. Concepts of nuclear reactors cooled with water at supercritical pressure were studied as early as the 1950s and 1960s in the USA and former USSR (Oka 2000; Wright and Patterson 1966; Bishop et al. 1962; Skvortsov and Feinberg 1961; Marchaterre and Petrick 1960; Supercritical pressure power reactor 1959). The main characteristics of the first concepts of SCWRs are listed in Table 4.1. After a 30-year interval, the idea of developing nuclear reactors cooled with supercritical water became attractive as the ultimate development path for water cooling. Several countries (Canada, Germany, Japan, Korea, Russia, USA, and others) have started R&D work in that direction. However, none of these concepts is expected to be implemented in practice before 2015–2020. The main objectives of using supercritical water in nuclear reactors are: 1. Increase the efficiency of modern nuclear power plants (NPPs) from 33%–35% to about 40%–45%; and 2. Decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US∕kW). Water is the most investigated fluid in near-critical and supercritical regions. All primary sources of heat-transfer experimental data of water flowing inside vertical circular tubes are listed in Table 5.1. Ranges of investigated parameters for selected experiments with water in circular tubes at supercritical pressures that are relevant to SCWR operating range are shown in Figure 5.1. Swenson et al. (1965) found that HTC has a peak when the film temperature is within the pseudocritical temperature range (Figures 5.2 and 5.3). This peak in HTC decreases with increase in pressure and heat flux. Vikhrev et al. (1971, 1967) conducted experiments in supercritical water flowing in a vertical tube (Figure 5.4). To be able to cover a wide range of bulk-fluid enthalpies (see Figure 5.4c,d), experiments were conducted in the same flow geometry, same mass flux and heat flux, but at various inlet temperatures (enthalpies). Later, these data were combined into one curve. However, this method is not the perfect one, because entrance and∕or exit (i.e., deteriorated heat transfer at the exit) effects do not allow matching properly two or several series in one plot (for details, see original figures in Vikhrev et al. (1967)). Therefore, it is important to perform supercritical heat-transfer experiments in one sufficiently long-heated test section. As mentioned previously, carbon dioxide has often been used as a supercritical working fluid in various thermodynamic cycles and heat exchangers and as a modeling fluid instead of supercritical water, because of its lower critical pressure and temperature. As mentioned previously, helium is also used as a coolant in some thermal units. However, a number of papers devoted to heat transfer to supercritical helium is significantly less than that devoted to water and carbon dioxide. Therefore, these papers are listed below just for completeness of the current review and reference purposes. Critical parameters and chemical formulas of the following fluids are listed in Table 2.1. Being influenced by the turbulent structure, it is expected that the heat transfer can be enhanced by having ribs, spacers, grids, etc. placed in the flow. Some sources of the primary data on heat-transfer enhancement at supercritical pressures in various fluids are listed in Table 9.1. The main idea of these experiments was to find the most effective ways to enhance supercritical heat transfer and to delay or to avoid the deteriorated heat-transfer regimes in various flow geometries. These studies were important to supercritical “steam” generators where heat fluxes are quite high. Therefore, supercritical water flowing in circular tubes was the primary choice for investigations. Some studies were performed in supercritical carbon dioxide as the modeling fluid. Also, due to limited space and weight, the supercritical heat-transfer enhancement in various hydrocarbon fluids and fuels was important for aircraft applications. Shiralkar and Griffith (1968) found that the twisted tape installed inside a bare tube cooled with supercritical carbon dioxide significantly improved the heat transfer (Figure 9.1). In this chapter, we examine and comment on the relevant features of experiments used to obtain heat-transfer and pressure-drop data in water and carbon dioxide based on experience at the IPPE (Kirillov et al. 2005), KPI (Pis'mennyy et al. 2005) and CRL (Pioro and Khartabil 2005; Pioro and Duffey 2003b). Additional information on this topic is given in Appendix E. Unfortunately, satisfactory analytical methods have not yet been developed due to the difficulty in dealing with the steep property variations, especially in turbulent flows and at high heat fluxes. Therefore, empirical generalized correlations based on experimental data are used for HTC calculations at supercritical pressures. McAdams (1942) proposed to use the Dittus and Boelter (1930) equation in the following form for forced convective heat transfer in turbulent flows and subcritical pressures (this statement is based on the recent study by Winterton (1998)): Nu b = 0.0243 Re b 0.8 Pr b 0.4 . Later, Equation (11.1) was also used for supercritical heat transfer. According to Schnurr et al. (1976) Equation (11.1) shows good agreement with the experimental data of supercritical water flowing inside circular tubes at pressure of 31 MPa and low heat fluxes. However, Equation (11.1) may give unrealistic results in some flow conditions, especially, near the critical and pseudocritical points, because it is sensitive to properties variations. In general, this classical equation was used intensively as the base for modified supercritical heat-transfer correlations. In general, the total pressure drop for forced convection flow inside a test section installed in the closed-loop system can be calculated according to the following expression: Δ p = ∑ Δ p f r + ∑ Δ p ℓ + ∑ Δ p a c + ∑ Δ p g , where Δ p is the total pressure drop, Pa. Δ p fr is the pressure drop due to frictional resistance (Pa), which defined as Δ p f r = ( ξ f r L D ρ u 2 2 ) = ( ξ f r L D G 2 2 ρ ) , where ξ fr is the frictional coefficient, which can be obtained from appropriate correlations for different flow geometries. For smooth circular tubes, ξ fr is as follows (Filonenko 1954): ξ f r = ( 1 ( 1.82 log 10 Re b − 1.64 ) 2 ) . Equation (12.3) is valid within a range of Re = 4.10 3 − 10 12 . Usually, thermophysical properties and the Reynolds number in Equations (12.2) and (12.3) are based on arithmetic average of inlet and outlet values. Δ p ℓ is the pressure drop due to local flow obstruction (Pa), which is defined as; Δ p ℓ = ( ξ ℓ ρ u 2 2 ) = ( ξ ℓ G 2 2 ρ ) , where ξ ℓ is the local resistance coefficient, which can be obtained from appropriate correlations for different flow obstructions. Unfortunately, satisfactory analytical methods have not yet been developed due to the difficulty in dealing with the steep property variations, especially in turbulent flows and at high heat fluxes. However, for completeness of the monograph it was decided to summarize the latest findings in analytical and numerical approaches in this chapter. For near-critical and supercritical pressures single-phase heat transfer, as well as for subcritical pressures two-phase heat transfer, flow oscillations can occur. These flow oscillations can significantly narrow the region with stable flow and affect heat transfer and reliability of heat-transfer equipment. For two-phase flows at subcritical pressures, the appearance of flow pulsations and thermo-acoustic oscillations are described in book by Gerliga and Scalozubov (1992). Also, Rohatgi and Duffey (1998) using homogenous equilibrium theory obtained a close-form solution for the stability region for parallel channels and derived the critical subcooling number, which is the lowest for unstable flow. Some conclusions from these two sources can be applicable for supercritical flows. Papers related to flow instability at supercritical pressures are listed below in the chronological order. Some general problems in the design and reliability of supercritical “steam” generators were discussed in papers by Styrikovich et al. (1967) and Rudyka et al. (1971). The following is a summary of the main findings of this study: • There is a plethora of data, empirical correlations and simple models for heat transfer and pressure drop published in the literature on the use of many supercritical fluids, mainly for simplified test sections. We have compiled and reviewed as much of the information as possible (in general, about 650 literature sources), and have provided all the key references and inter-compared the experimental and theoretical approaches. The data largely cover all the ranges of interest, but of course for mainly commercial reasons some design specific information is missing from the open papers and reports. • There are hundreds of fossil power plants in the world using supercritical conditions (thermal parameters: water pressure of up to 25 – 30 MPa, turbine inlet temperatures of up to 625°C (but mainly lower than 600°C) and power output of up to 1400 MW e ), which have been successfully operated for many years. Their main advantage is high thermal efficiency of up to 45% – 53%. The demonstrated experience in their design and operation is very helpful for current developments in fossil-fired units and in nuclear-powered reactors cooled with supercritical water and provides useful benchmark data. • After a 30-year hiatus, because of the fossil experience and the need to improve the overall cycle thermal efficiency, the idea of developing nuclear reactors cooled with supercritical water became attractive again, and several countries (Canada, Germany, Japan, Russia, and the USA) have started to work in that direction. However, none of these concepts is expected to be implemented in practice before 2015 – 2020. This back matter contains appendix A-E, references and index