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The main criteria used in NEI 18–04 to define SSCs as risk-significant include (1) the SSC is required to keep all LBEs within the F–C target, and (2) the total frequency with the SSC failed exceeds 1% of the limit for at least one of the three cumulative risk metrics used for evaluating the integrated plant risk. The first one is a reasonable criterion in determining the risk significant SSCs. However, the second criterion may not be adequate to serve the purpose of determining the risk significance of SSCs. In the second criterion, the cumulative risk metric values representing the integrated plant risk (less the preventive and mitigative effects of the SSC being evaluated) are compared to a risk limit that represents a very small contribution to the overall integrated plant risk, which corresponds appropriately to the contributions from individual SSCs. The easiest approach to redefine the NEI 18-04 definition of risk-significant SSCs in relation to the integrated plant risk metrics is to compare the difference, between the risk metric value calculated with the SSC failed and the risk metric value calculated with the SSC credited, with 1% of the risk limit established for the integrated plant risk metrics.

Nuclear security has been emphasized to ensure the safety of the environment and humans, as well as to protect nuclear materials and facilities from malicious attacks. With increasing utilization of nuclear energy and emerging potential threats, there has been a renewed focus on nuclear security. Korea has made efforts to enhance the regulatory oversight processes, both for general and specific legislative systems. While Korea has demonstrated effective nuclear security activities, continuous efforts are necessary to maintain a high level of security and to improve regulatory efficiency in alignment with international standards.In this study, the comprehensive regulatory oversight framework for the security of Korean nuclear facilities has been investigated. For reference, the U.S. regulatory oversight frameworks for nuclear facilities, with a focus on nuclear security, and the motivations of changes in regulatory oversight framework have been identified. By comparing these regulatory programs and frameworks, insights and considerations for enhancing nuclear security regulations have been identified. A comprehensive security inspection program tailored for the Korean regulatory oversight framework has been proposed, and has been preliminarily applied to hypothetical conditions for further discussion.

A lack of consideration for maintainability at the planning/design stage of a construction project often results in inadequate decision-making and selection of design alternatives, which will likely lead to unproductive operational regimes riddled with frequent defects. This paper attempts to anchor the quantitative risk analysis technique, Bayesian networks, as apt in predicting future operational defects based on the initial design alternatives by looking at specific case studies in building basement systems. Therefore, this paper aims to develop a comprehensive risk-informed, reliability-driven decision-making system. Central to this process is the implementation of a setup wherein a Bayesian Belief Network (BBN) operates as a powerful reliability engine. Concurrently, a risk matrix serves as a vital tool for facilitating risk-informed decision-making. This approach empowers stakeholders to make informed decisions regarding design choices while keeping maintainability at the forefront. It allows for maintainability to be practically used as an effective tool in buildings to lower the lifecycle cost. Furthermore, it is envisaged that this approach can be adopted in the construction industry globally for the prediction and diagnosis of defects, thereby facilitating risk-informed decision support. This study provides significant contributions to the building industry by providing a novel approach to a basement defect analysis, encouraging interdisciplinary collaboration among construction stakeholders, and assisting facility engineers and managers in properly determining maintenance requirements. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The risk-informed safety margin is a new safety concept of nuclear power industry in recent ten years. The calculation framework and the quantification technology under the Monte Carlo method of the riskinformed safety margin were studied. The riskinformed safety margin quantification technology of the station blackout (SBO) accident of nuclear power plant was focused on with the Monte Carlo sampling method. The simplified thermal hydraulic model of nuclear power plant was established with reference parameters of SBO accident in the simplified pressurized water reactor (PWR) model in the INL’s (Idaho National Laboratory) RISMC research. And the calculation process of the riskinformed safety margin based on Monte Carlo sampling method was determined. The number of scenarios generated by Monte Carlo sampling was essential to ensure the acceptability of the accuracy of the calculation results, and the accuracy of estimated value of probability safety margin would be higher if we establish more scenarios. The sampling times estimation method and the reliability calculation method of Monte Carlo were referenced, and a quantitative method of riskinformed safety margin with Monte Carlo sampling was proposed. The sampling times estimation method under the given error was given. According to the uncertainty calculation method of the riskinformed safety margin with Monte Carlo method and the estimation process of sampling times in Monte Carlo, the sampling times were determined by relative error and absolute error respectively. The results show that absolute error is equal to 001 with the 4 706 sampling times, and relative error is equal to 5% with the 9 604 sampling times in the riskinformed safety margin in SBO accident. The mean value and standard deviation of the failure probability of nuclear fuel cladding were calculated with these two sampling times. At this stage with the usual calculation of resource allocation, it is suggested that when the average failure probability of nuclear fuel cladding is larger than 01, the sampling times with 5% relative error should be selected. And when the mean value of failure probability of nuclear fuel cladding is equal to or lesser than 01, the sampling times with relative error equal to 10% could be used. The change of the mean value of failure probability of nuclear fuel cladding under the Latin hypercube sampling and random sampling method was studied. The relationship of the mean value with the recovery time of emergency diesel engine under different distributions was researched too. The calculation results indicate that the different sampling methods and different normal distributions have significant effects on mean value and standard deviation of the failure probability of the nuclear fuel cladding.

A lack of consideration for maintainability at the planning/design stage of a construction project often results in inadequate decision-making and selection of design alternatives, which will likely lead to unproductive operational regimes riddled with frequent defects. This paper attempts to anchor the quantitative risk analysis technique, Bayesian networks, as apt in predicting future operational defects based on the initial design alternatives by looking at specific case studies in building basement systems. Therefore, this paper aims to develop a comprehensive risk-informed, reliability-driven decision-making system. Central to this process is the implementation of a setup wherein a Bayesian Belief Network (BBN) operates as a powerful reliability engine. Concurrently, a risk matrix serves as a vital tool for facilitating risk-informed decision-making. This approach empowers stakeholders to make informed decisions regarding design choices while keeping maintainability at the forefront. It allows for maintainability to be practically used as an effective tool in buildings to lower the lifecycle cost. Furthermore, it is envisaged that this approach can be adopted in the construction industry globally for the prediction and diagnosis of defects, thereby facilitating risk-informed decision support. This study provides significant contributions to the building industry by providing a novel approach to a basement defect analysis, encouraging interdisciplinary collaboration among construction stakeholders, and assisting facility engineers and managers in properly determining maintenance requirements.

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE). The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300-MW(thermal), sodium-cooled, metallic-fueled, pool-type fast reactor has been led by U.S. national laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. To facilitate risk-informed design and authorization activities during the conceptual development phase, a conceptual design probabilistic risk assessment (PRA) was performed for the VTR. This paper provides an overview of the development of the VTR conceptual design PRA, including key DOE and industry standards and the PRA analysis approach and structure. In addition, the results of the VTR conceptual design PRA are provided, which include its use within authorization documentation and design decisions, along with important lessons learned during the process. The work reported in the paper is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary. [ABSTRACT FROM AUTHOR]

The terms "prevention" and "mitigation" are often used to define safety functions in nuclear power plants and other complex engineered systems. For existing light water reactor (LWR) nuclear plants, these terms are typically used in the context of preventing core damage and mitigating the consequences of a severe accident. Balancing the reliance on prevention and mitigation is often stated as an important principle of defense-in-depth. To apply these terms to formulate design requirements for advanced reactors, it is necessary to define specifically what is to be prevented and mitigated because the concept of core damage is not generally applicable to advanced non-LWR concepts and designs. Advanced non-LWR reactor concepts use different fuels, moderators, coolants and different strategies for achieving the functional containment of radionuclides. The adoption of the small modular reactor concept also leads to plants with many small reactors. This paper discusses a risk-informed and performance-based definition of prevention and mitigation that can be applied to any reactor concept or design. It was developed as part of the Licensing Modernization Project (LMP) which has introduced a new approach to developing a safety case to support design and licensing. A new way of thinking about prevention and mitigation was needed to formulate advanced reactor design and special treatment requirements for structures, systems and components (SSCs) that could be applied to a wide diversity of designs. The LMP approach discussed in this paper includes criteria to identify safety significant SSC prevention and mitigation functions and uses SSC reliability and capability targets to inform the selection of design and special treatment requirements and to support the evaluation of defense-in-depth adequacy. These prevention and mitigation functions are defined in the context of event sequences that comprise the licensing basis events that anchor the safety case for the reactor operating license. [ABSTRACT FROM AUTHOR]

Regulators around the world have devised practical methods for evaluating the safety of nuclear power plants (NPPs) and increasing public awareness of plant safety. One essential approach is the regulatory safety-related inspection, which plays a crucial role in confirming the safety of power plants to prevent and mitigate accidents. In recent years, many developed countries have adopted a risk-informed methodology, which, similar to the conventional deterministic approach, prioritizes various aspects of regulatory inspection which relies on PSA insights for decision-making, rather than being rooted in a deterministic approach. In the United Arab Emirates, where three nuclear units are in operation, the regulatory inspection focuses on reducing the likelihood of incidents through a defense-in-depth (DiD) concept and the insights gained from risk analysis. To strengthen accident prevention oversight, both deterministic and probabilistic approaches are utilized, including probabilistic risk assessments (PRA) aimed at identifying safety improvements. This study aims to develop a preliminary safety inspection methodology using AIMS-PSA as an analysis tool. It involves constructing event trees and fault trees to understand how to secure the success path of Loss of Coolant Accidents (LOCAs) transients, block potential failure paths, and assess their impact on Core Damage Frequency (CDF) by analyzing the generated cut sets. This study will offer quantifiable insights into key risk factors, improving our capacity to anticipate and prevent accidents, maintain high reliability, and conduct effective risk assessments for the continued operability of safety systems. While previous investigations have explored nuclear accident frequencies, often with limitations, this research aims to provide a more comprehensive perspective on nuclear power plant safety, addressing a wide range of accident scenarios and their implications. • Regulatory safety inspections are crucial for verifying nuclear power plant safety. • The UAE's focus on minimizing incident probability using a defense-in-depth approach. • AIMS-PSA is used to analyze LOCA transient success and failure paths. • The study identifies core contributors to risk and aids in accident anticipation. • The methodology enhances safety, reliability, and long-term accident mitigation. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Risk-informed, performance-based (RIPB) methods have progressed to the point where high-level guidance can be used to augment traditional, deterministic, nuclear safety design practices in areas important to nuclear reactor safety. This paper describes an approach for augmenting the traditional defense-in-depth (DID) qualitative approach with quantitative risk information from a plant-specific probabilistic risk assessment (PRA) in a way that is structured, can be applied on a consistent basis, and allows for clear acceptance criteria. Adding performance-based targets that should be achieved is expected to result in safer and more economical plant designs. Evaluations of DID can be conducted throughout the design process as well as in support of design certification and operating license applications to identify where defense protections could be enhanced or relaxed. Consistent with the United States Nuclear Regulatory Commission's policy statement encouraging greater use of PRA to improve safety decision making and regulatory efficiency, this scenario-based DID method can be used to evaluate changes and overall plant design as part of the normal design control process. Although the RIPB method presented in this paper was developed for application to advanced passive light water reactor designs, the metrics could be tailored to other reactor designs. This risk-informed approach to DID helps to ensure that public and worker risk insights are integrated into the design process holistically. [ABSTRACT FROM AUTHOR]

Summary: This paper proposed a new importance assessment method and new importance measure to perform risk‐informed system, structure, and components (SSCs) categorization task, regarding the dynamic of SSCs' importance measures caused by varied configurations in nuclear power plants (NPPs). The new method consists of 3 importance measures, Fussell Vesely, risk achievement worth, and return to service worth, and an integrated time ratio factor is introduced to get an SSCs ranking, which could handle more realistic configurations in NPP operation. The proposed method has been used in CANDU feedwater system SSCs categorization, supported by RiskA and RiskAngel software, and a different SSCs categorization result compared with NEI method was given. The difference of categorization result derived from 2 methods was discussed in this paper, and the effectiveness has been demonstrated by the NPP's real operation statistics. [ABSTRACT FROM AUTHOR]

The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the K Ia fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the K IC fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan. [ABSTRACT FROM AUTHOR]

The evaluation of methodology for large break loss of coolant accident (LBLOCA) licensing analysis involves two kinds of methodologies, namely deterministic methodology and risk-informed methodology. According to the existing 10 CFR50.46, LBLOCA is one of the most essential design-basis accidents and a deterministic methodology shall be applied to perform LBLOCA analysis based on a so-called surrogate sequence. Without considering how low this sequence occurrence probability is, this surrogate sequence satisfies all the required licensing assumptions. However, in the to-be-issued 10 CFR 50.46a, the LBLOCA can be categorized as an accident beyond design basis and the margin of peak cladding temperature (PCT) shall be evaluated in a risk-informed manner. Regarding the deterministic LBLOCA methodologies, the PCT margin quantified by the traditional appendix K methodology with conservative appendix K models and bounding plant parameters can be enlarged by 41.5 K with the deterministic-realistic hybrid methodology (DRHM), in which realistic plant parameters and statistical uncertainty analysis are utilized. It was noted that the PCT margin released by realistic plant parameters can be about 1/4–1/5 of the margin generated by a full-scoped BEPU methodology with both best-estimate models and realistic plant parameters. Furthermore, by relaxing the licensing sequence assumptions, an additional LBLOCA PCT margin of 44.3–62.2 K can be released from the deterministic approaches by applying the risk-informed safety margin characterization (RISMC) methodology. It was also observed from the risk-informed LBLOCA analysis, to have a cumulated occurrence probability over 99% in the load spectrum, the third sequence by order of ascending PCT was referred and the associated sequence occurrence probability of this sequence is 5.07 * 10 −3 . While in the deterministic methodology, the occurrence probability of the traditional surrogate or licensing sequence is only 5.46 * 10 −5 . [ABSTRACT FROM AUTHOR]

This paper studied the requirements of risk monitor software and its applications as a plant specific risk monitor, which supports risk-informed configuration risk management for the two CANDU 6 units at the Third Qinshan nuclear power plant (TQNPP) in China. It also describes the regulatory prospective on risk-informed Probabilistic Safety Assessment (PSA) applications and the use of risk monitor at operating nuclear power plants, high level technical and functional requirements for the development of CANDU specific risk monitor software, and future development trends. [ABSTRACT FROM AUTHOR]

This paper presents an approach and an example of application for the evaluation of risk impact of changes to Completion Times within the License Basis of a Nuclear Power Plant based on the use of the Probabilistic Risk Assessment addressing identification, treatment and analysis of uncertainties in an integrated manner. It allows full development of a three tired approach (Tier 1-3) following the principles of the risk-informed decision-making accounting for uncertainties as proposed by many regulators. Completion Time is the maximum outage time a safety related equipment is allowed to be down, e.g. for corrective maintenance, which is established within the Limiting Conditions for Operation included into Technical Specifications for operation of a Nuclear Power Plant. The case study focuses on a Completion Time change of the Accumulators System of a Nuclear Power Plant using a level 1 PRA. It focuses on several sources of model and parameter uncertainties. The results obtained show the risk impact of the proposed CT change including both types of epistemic uncertainties is small as compared with current safety goals of concern to Tier 1. However, what concerns to Tier 2 and 3, the results obtained show how the use of some traditional and uncertainty importance measures helps in identifying high risky configurations that should be avoided in NPP technical specifications no matter the duration of CT (Tier 2), and other configurations that could take part of a configuration risk management program (Tier 3). [ABSTRACT FROM AUTHOR]

Abstract: The paper presents the application of a new risk-informed methodology for the identification of the Emergency Management Requirements (EMR) to a Generation II, Large size Reactor and a Generation III+ Small Modular Reactor. The results obtained in this test case demonstrate that the actual EMR is conservative, as expected, for the GenII reactor, while the new methodology could be applied for the definition of EMRs for the new generation Nuclear Power Plants, with a possible reduction of the emergency area without loss of safety level. By adopting both probabilistic and deterministic approaches, the study addresses possible accidents and corresponding release scenarios for the two types of reactor, calculates the areas where the accidents have an impact on the population and defines the new EMR considering the health effects on the population. [Copyright &y& Elsevier]

• Multi-objective route planning indicators are established for radioactive material road transportation considering radiological risk cost, transportation time cost, and economical cost. • A novel integrated probabilistic safety assessment (PSA) methodology framework is proposed for the radiological risk cost estimation. • To obtain the risk, the accident sequence-based probability analysis model and the PEHR comprehensive severity index-based radiological consequence evaluation model are established. • The proposed RICPP methodology is beneficial for selecting a safer and more economical route for radioactive materials road transportation. The optimization of transportation routes of radioactive materials (RADMAT) is recognized as one of the most important measures to reduce the radioactive risks and costs during transportation. However, the route optimization indicators have not been fully considered and quantified in current researches. Here, we propose a Risk-Informed based Comprehensive Path-Planning method (RICPP) for RADMAT Road Transportation (RMRT) route optimization by considering radiological risk cost, time cost, and economical cost. Among which, the ET/FT based Probabilistic Safety Assessment (PSA) model is established to quantitatively assess the accident occurrence probability, and then the PEHR (personnel, environment, hazard target, and rescue force) comprehensive severity indexes considering vulnerability and resilience are established and quantified to characterize the accident radiological consequence. Finally, the gray relation analysis (GRA) model is employed to select an optimal path considering the multi-objective route planning indicators. The case study results show that the obtained path could reasonably reduce risks and costs during the transportation process, which demonstrates the effectiveness of the proposed method. The proposed RICPP methodology is beneficial for selecting a safer and more economical route for radioactive materials road transportation. [ABSTRACT FROM AUTHOR]

Individuals make decisions every day in group contexts which vary in size, structure, and purpose. The US Department of Defense (DoD) is a large organization composed of many groups, and like many organizations, it has a vested interest in improving the performance of its affiliated groups, especially as it concerns risk-informed decision-making. This article discusses current foibles and considerations for decision-making in DoD groups as identified through a workshop with experts in risk-informed decision-making, cognitive science, and military operations. Experts noted that terms associated with risk-informed decision-making were often misconstrued, that formal decision-making frameworks are underutilized, and that many considerations should be taken into account when attempting to improve decision-making performance. [ABSTRACT FROM AUTHOR]

Performance-based building codes are being developed and promulgated around the world. Concurrently, performance-based analysis and design approaches are being used in a number of disciplines, including structural, mechanical, and fire protection engineering. The performance-based building regulatory and design environment promises great opportunities for engineers and designers to innovate and to apply analytical tools and methods to design safe, efficient, cost-effective, and aesthetically pleasing buildings. However, for regulators and enforcement officials, performance-based approaches are often met with skepticism and concern, as the desired performance is not always well defined and agreed, the perceived certainty associated with compliance with prescriptive design requirements is no longer assured, and there is concern that the data, tools, and methods - necessary to assure that performance-based designed buildings achieve the levels of performance and risk deemed tolerable to society - are lacking. To address these concerns, risk-informed performance-based approaches are being explored, with the aim to better identify and connect tolerable levels of risk, performance expectations, and design criteria for different aspects of building design. Risk-informed performance-based approaches being considered in Australia, New Zealand, and the USA are discussed. [ABSTRACT FROM AUTHOR]

Abstract: This paper shows an efficient order for calculating importance measures and develops several new measures related to fault diagnostics, system failure intensity, system failure count, and configuration control. Definitions and relationships are extended to certain non-coherent systems and to models with mutually explicit events. Useful interpretations and applications are pointed out, and many roles of the Birnbaum importance are highlighted. Another important topic is the accuracy of various alternative methods used for quantification of accident sequence probabilities when negations or success branches of event trees are involved. Finally, the role of truncation errors is described, and criteria are developed for selecting truncation limits and cut-off errors so that importance measures can be estimated reliably and risk-informed decision making is robust, without unreasonable conservatism and without unwarranted optimism. [Copyright &y& Elsevier]

To meet London Protocol requirements, South Korea is preparing to reduce the need for disposal of dredged material at sea. The new requirements controlling ocean disposal of dredged material pose significant challenges to the South Korean government, because the previous practice of offshore disposal of contaminated dredged material is no longer permitted. Hence, other alternatives for treating and disposing of contaminated dredged material are being evaluated and selected for implementation. A new management and decision approach is therefore needed for regulators and implementers to show what information and what decision-making processes were used to make the decision, to increase administrative transparency for such projects in the public domain. To address this need, an iterative approach was developed for dredged material management that includes the essential elements of process, people, and tools needed for successful environmental decision making. The approach has 6 steps: problem definition, developing objectives and criteria, identifying alternatives, performing the evaluation, comparing alternatives, and selecting the preferred alternative. The primary objective of the approach is to provide a systematic means of exploring contaminated dredged material management alternatives in South Korea using criteria that integrate risk with economic and stakeholder value information. The approach incorporates the desired decision-making attributes of transparency, comparative analysis, and inclusion of public input. Although it was developed for South Korea, the approach can be applied in any situation in which dredged material management alternatives are being considered to manage contaminated sediment risks. [ABSTRACT FROM AUTHOR]

Abstract: The permitting process for hydrogen fueling stations requires demonstration that the proposed facility meets certain safety requirements. Currently, many permitting authorities rely on compliance with well-known codes and standards as evidence of a safe design. To ensure that a hydrogen facility is indeed safe, the code and standard requirements should be identified using a risk-informed process that utilizes an acceptable level of risk. This paper describes an approach for risk-informing the permitting process for hydrogen fueling stations that relies primarily on the establishment of risk-informed codes and standards. Using accepted Quantitative Risk Assessment (QRA) techniques and the established risk criteria, the minimum code and standard requirements necessary to ensure the safe operation of hydrogen facilities can be identified. To facilitate consistent risk-informed approaches, the participants in the International Energy Agency (IEA) Task 19 on hydrogen safety are working to identify acceptable risk criteria, QRA models, and supporting data. [Copyright &y& Elsevier]

Abstract: Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant''s availability, decreases dose rate on the plant''s staff and also reduces cost on plant''s operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant''s licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant''s safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant''s management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs’ and AOVs’ safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee. [Copyright &y& Elsevier]

Abstract: The Idaho National Laboratory is a primary developer of probabilistic risk and reliability analysis (PRRA) tools, dating back over 35 years. Evolving from mainframe-based software, the current state-of-the-practice has led to the creation of the SAPHIRE software. Currently, agencies such as the Nuclear Regulatory Commission, the National Aeronautics and Aerospace Agency, the Department of Energy, and the Department of Defense use version 7 of the SAPHIRE software for many of their risk-informed activities. In order to better understand and appreciate the power of software as part of risk-informed applications, we need to recall that our current analysis methods and solution methods have built upon pioneering work done 30–40 years ago. We contrast this work with the current capabilities in the SAPHIRE analysis package. As part of this discussion, we provide information for both the typical features and special analysis capabilities, which are available. We also present the application and results typically found with state-of-the-practice PRRA models. By providing both a high-level and detailed look at the SAPHIRE software, we give a snapshot in time for the current use of software tools in a risk-informed decision arena. [Copyright &y& Elsevier]

The Advisory Committee on Nuclear Waste advises the Nuclear Regulatory Commission on public health and safety issues in relation to nuclear materials and nuclear wastes. The advice includes safety and environmental protection issues in low- and high-level waste, decontamination and decommissioning of facilities, transportation of nuclear materials, and safeguards and security. High-level waste issues have been principally related to a potential repository at Yucca Mountain, Nevada. Other issues include decommissioning, public outreach, low-level waste, and methods of safety analysis. A principal activity of the Committee has been assisting the Nuclear Regulatory Commission in its transition to a risk-informed regulatory practice. [ABSTRACT FROM AUTHOR]

A risk-informed safety significance categorization (RISSC) is to categorize structures, systems, or components (SSCs) of a nuclear power plant (NPP) into two or more groups, according to their safety significance using both probabilistic and deterministic insights. In the conventional methods for the RISSC, the SSCs are quantitatively categorized according to their importance measures for the initial categorization. The final decisions (categorizations) of SSCs, however, are qualitatively made by an expert panel through discussions and adjustments of opinions by using the probabilistic insights compiled in the initial categorization process and combining the probabilistic insights with the deterministic insights. Therefore, owing to the qualitative and linear decision-making process, the conventional methods have the demerits as follows: (1) they are very costly in terms of time and labor, (2) it is not easy to reach the final decision, when the opinions of the experts are in conflict and (3) they have an overlapping process due to the linear paradigm (the categorization is performed twice—first, by the engineers who propose the method, and second, by the expert panel). In this work, a method for RISSC using the analytic hierarchy process (AHP) and bayesian belief networks (BBN) is proposed to overcome the demerits of the conventional methods and to effectively arrive at a final decision (or categorization). By using the AHP and BBN, the expert panel takes part in the early stage of the categorization (that is, the quantification process) and the safety significance based on both probabilistic and deterministic insights is quantified. According to that safety significance, SSCs are quantitatively categorized into three categories such as high safety significant category (Hi), potentially safety significant category (Po), or low safety significant category (Lo). The proposed method was applied to the components such as CC-V073, CV-V530, and SI-V644 in Ulchin Unit 3 NPP in South Korea. The expert panel consisted of two probabilistic safety assessments (PSA) experts and one system design expert. Before categorizing the components, the design basis functions, simplified P and IDs, and the Fussell-Vesely (FV) importance and the Risk Achievement Worth (RAW) of the PSA were prepared for the experts'' evaluations. By using this method, we could categorize the components quantitatively on the basis of experts'' knowledge and experience in an early stage. [Copyright &y& Elsevier]

The recent overall improvement in key operational safety indices in the United States, combined with ‘risk-informed, performance-based’ regulation by the US Nuclear Regulatory Commission (NRC), has indicated that ‘safety goals’ are indispensable; thereby both licensee and regulator can share common objectives and common indicators of safety performance. Recognizing these, the author proposes a new concept of safety goals11See also ‘A New Approach to Reactor Safety Goals in the Framework of INES,’ another article in this issue. to facilitate engineering application, while removing some of the uncertainties often encountered in implementing the safety goals, by extending a framework of the International Nuclear Event Scales (INES) being widely used in the world. In this article, safety goals are characterized from a point of view of nuclear regulation by oversight, as established by the US NRC. This is a new tendency of nuclear regulation to motivate initiatives of licensees to improve safety and operational performance and to minimize potential nuclear risks, without the regulatory side specifying how the specific safety requirements should be met. Whereas in the ‘compliance-based regulation,’ which is a more widely used approach of nuclear regulation in many countries, detailed prescriptive safety requirements are specified to enforce the licensees to strictly follow them. The author observes, through the past experience of the US NRC, the latter approach has a basic limitation in improving total safety of nuclear facilities, and supports the new direction to be taken more widely in the nuclear community. [Copyright &y& Elsevier]

This work describes the use of fuzzy logic methodology to estimate inservice inspection (ISI) priorities for nuclear components with Adaptive Neural Fuzzy Inference system (ANFIS). To estimate ISI priorities for components the methodology uses probabilities of component failures and the core damage frequency of each component failure. The specific systems addressed in this work are the auxiliary feedwater, low-pressure injection, and reactor coolant systems selected from the Surry Nuclear Power Station Unit 1 (Surry-1). For illustration some examples were developed and their results, when compared with reference work lead to important conclusions. [Copyright &y& Elsevier]

The mainstream deterministic and probabilistic methods for safety analysis of the nuclear power plant (NPP) are the best estimate plus uncertainty (BEPU) analysis and probabilistic safety analysis (PSA), respectively. With the development of techniques, a more "realistic" simulation of the NPP has become a long-term goal to improve the economic efficiency. Nowadays, the combination of BEPU and PSA, also known as the risk-informed uncertainty method, is under development. In this paper, the cold leg double-ended break accident of a typical pressurized water reactor was selected to be the target scenario for a preliminary study of the combined method. A failure analysis of the emergency core cooling system based on PSA was carried out first, and the failure probability of each event sequence was determined. Then the conditional core damage probability of each event sequence was evaluated through BEPU analysis, and the core damage frequency of the reactor was finally determined. [ABSTRACT FROM AUTHOR]

• Risk informed methodology for system-level validation. • Identification of events critical for improving system-level validation. • Propose consistency index to ensure system-level validation is complete. • Adequacy of a simulation code for an application is characterized by maturity level. In recent years, the U.S. Nuclear Regulatory Commission (USNRC) and the International Atomic Energy Agency (IAEA) have developed methodologies to assess the vulnerabilities of nuclear plants against site specific extreme hazards. In many cases, advanced simulation tools are being considered to simulate multi-physics, multi-scale phenomena and to evaluate vulnerability of nuclear facilities. The credibility of advanced simulation tools is assessed based on a formal verification, validation, and uncertainty quantification procedure. One of the key limitations in validation is the lack of relevant experimental data at system-level. This limitation leads to a decrease in the confidence of system-level risk predictions. Therefore, a robust validation framework is needed to formalize the confidence in predictive capability of advanced simulation results. This study enhances the existing risk informed validation methodology, originally proposed by Kwag et al. [1] and Bodda et al. [2], by developing additional attributes and a new set of validation indicies for a complete and wider applicability of the framework. In this manuscript, the methodology to identify the critical path that leads to the system-level failure is illustrated. The overall validation is checked for completeness and consistency by comparing the critical path for both the system-level simulation and experimental models. The applicability of the code for an intended application is represented in terms of various maturity levels and helps in the process of decision making. [ABSTRACT FROM AUTHOR]

• Risk management technical specifications framework. • Harmonisation of surveillance requirements and maintenance. • RAM and risk models in a context of equipment ageing and obsolescence. • Test and maintenance optimisation in the context of RMTS philosophy. • Usefully application for the renewal of the licence for operation of the plant. This paper contains the fundamentals and a comparison of different alternatives in the context of risk-informed surveillance requirements that consider the full effects of implementing the maintenance rule. A case study is included to demonstrate the performance of the different alternatives, which focus on a motor-operated valve and make use of an Ageing PRA to quantify the effect of component ageing, test and maintenance effectiveness on equipment RAM and its impact on risk. The results show that the alternative that simultaneously harmonises surveillance requirements and maintenance activities provides the best results in terms of RAM and risk in a context of equipment ageing and obsolescence. However, they also show that measures other than simply re-adjusting surveillance testing and maintenance intervals should be explored in case of technical obsolescence, as this has a strong impact on maintenance effectiveness, which may require, for example, an obsolescence management program to be applied. [ABSTRACT FROM AUTHOR]