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[en] Highlights: • A safety culture framework and a quantitative methodology to assess safety culture were proposed. • The relation among Norm system, Safety Management System and worker's awareness was established. • Safety culture probability at NPPs was updated by collecting actual organizational data. • Vulnerable areas and the relationship between safety culture and human error were confirmed. - Abstract: For a long time, safety has been recognized as a top priority in high-reliability industries such as aviation and nuclear power plants (NPPs). Establishing a safety culture requires a number of actions to enhance safety, one of which is changing the safety culture awareness of workers. The concept of safety culture in the nuclear power domain was established in the International Atomic Energy Agency (IAEA) safety series, wherein the importance of employee attitudes for maintaining organizational safety was emphasized. Safety culture assessment is a critical step in the process of enhancing safety culture. In this respect, assessment is focused on measuring the level of safety culture in an organization, and improving any weakness in the organization. However, many continue to think that the concept of safety culture is abstract and unclear. In addition, the results of safety culture assessments are mostly subjective and qualitative. Given the current situation, this paper suggests a quantitative methodology for safety culture assessments based on a Bayesian network. A proposed safety culture framework for NPPs would include the following: (1) a norm system, (2) a safety management system, (3) safety culture awareness of worker, and (4) Worker behavior. The level of safety culture awareness of workers at NPPs was reasoned through the proposed methodology. Then, areas of the organization that were vulnerable in terms of safety culture were derived by analyzing observational evidence. We also confirmed that the frequency of events involving human error decreases when the level of safety culture is high. It is anticipated that the causality between the safety culture awareness of worker and the state of safety at NPPs can be verified using the proposed methodology.
[en] For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must be presented at the Regulatory Authority with the necessary confidence on the models used to describe the plant safety behavior. In principle, this requires the repetition of a large number of model runs to account for the uncertainties inherent in the model description of the true plant behavior. The present paper propounds the use of bootstrapped Artificial Neural Networks (ANNs) for performing the numerous model output calculations needed for estimating safety margins with appropriate confidence intervals. Account is given both to the uncertainties inherent in the plant model and to those introduced by the ANN regression models used for performing the repeated safety parameter evaluations. The proposed framework of analysis is first illustrated with reference to a simple analytical model and then to the estimation of the safety margin on the maximum fuel cladding temperature reached during a complete group distribution header blockage scenario in a RBMK-1500 nuclear reactor. The results are compared with those obtained by a traditional parametric approach
[en] To meet the energy requirement for a remote and isolated region, a concept of supercritical CO2 (S-CO2) cooled micro modular reactor (MMR) has been developed, which is easy to transport by fully modularizing the nuclear power system. From the previous work, only the on-design performances of MMR were presented. However, since the nuclear system needs to meet high safety standard, the design can be finalized only after the evaluation of the nuclear system response under multiple transient conditions. Such transient conditions may include part load operation and postulated accidents. The system response can be evaluated only with a high fidelity computer simulation, since it is almost impossible to construct a nuclear facility to test at such conditions. The simulation is performed to guarantee its autonomous operability and safety. Unfortunately, a system analysis computer code available for the S-CO2 cycle analysis is yet to be developed for the MMR safety evaluation. Description of a S-CO2 system analysis code platform is first presented in this paper. After that, the developed code is validated with experimental data. The steady state of MMR is first modeled with the code, which is followed by simulations of part load operation and postulated accidents of MMR. All simulation results show that the current design of MMR has the ability to maintain its safety and structural integrity even under extreme conditions to protect the public from radiation hazard at all times.
[en] This paper deals with the structure of public attitude towards nuclear power plants in Korea. Special emphasis is given to the issues of public acceptance in relation to perceived benefits, perceived risk, judged safety, and safety satisfaction. The national survey data of 1995 by the Korea Institute of Nuclear Safety is analyzed with a latent class model and logistic regression. The latent class model is used to construct benefit and risk factors. With these factors and safety-related variables, a public attitude model is developed by logistic regression which enables the relationships between national or local acceptance of nuclear power and explanatory variables to be quantified. The results show that the attitude structure is somewhat different by gender. Subjectively perceived risk is found to be the most influential factor for local acceptance. The odds of local acceptance with the best risk perception is about 16 times the odds with the worst perception for males and about 7.7 times for females. From the results of this analysis, it is clear that subjective satisfaction with nuclear safety is a more important factor for explaining public acceptance rather than judgment of it. These important findings should be reflected in the public acceptance improvement strategy for the nuclear power program
[en] Highlights: • The choice and justification of operational software and systems are reviewed. • The standards IEC 61513:2011, ISO/IEC 23360-1:2006 and ISO/IEC 27032:2012 are analysed. • The question whether the Linux operational system conforms to the demands of the above standards is discussed. - Abstract: In this paper, we discuss problems of selecting and justifying operational software, especially operating systems. Operating systems must meet the requirements of international organisations (e.g., IEAE, IEC, ISO). The most important standards (IEC 61513:2011, ISO/IEC 23360-1:2006 and ISO/IEC 27032:2012) were analysed. Furthermore, the issue of conformance one of the most widespread operating system Linux to the requirements of the standards was analysed
[en] In this paper we address the importance of including the consideration of revenue loss into the safety analysis as well as system optimisation and modify the traditional Life Cycle Cost (LCC) into Life Cycle Revenue Loss (LCRL) as the criterion of optimisation and a quantitative assessment of the consequence of un-wished ebents, such as system unavailability. Through the Monte Carlo simulation technique and a simple scenario of decision making in a bidding process, we demonstrate the feasibility of our new LCRL model
[en] International regulations for nuclear power plants strictly prescribe the design requirements for local impact loads, such as aircraft engine impact, and internal and external missile impact. However, the local impact characteristics of Steel-plate Concrete (SC) walls are not easy to evaluate precisely because the dynamic impact behavior of SC walls which include external steel plate, internal concrete, tie-bars, and studs, is so complex. In this study, dynamic impact characteristics of SC walls subjected to local missile impact load are investigated via actual high-speed impact test and numerical simulation. Three velocity checkout tests and four SC wall tests were performed at the Energetic Materials Research and Testing Center (EMRTC) site in the USA. Initial and residual velocity of the missile, strain and acceleration of the back plate, local failure mode (penetration, bulging, splitting and perforation) and deformation size, etc. were measured to study the local behavior of the specimen using high speed cameras and various other instrumentation devices. In addition, a more advanced and applicable numerical simulation method using the finite element (FE) method is proposed and verified by the experimental results. Finally, the experimental results are compared with the local failure evaluation formula for SC walls recently proposed, and future research directions for the development of a refined design method for SC walls are reviewed.
[en] Highlights: •We proposed a Probabilistic Safety Culture Healthiness Evaluation Method. •Positive relationship between the ‘success’ states of NSC and performance was shown. •The state probability profile showed a unique ratio regardless of the scenarios. •Cutset analysis provided not only root causes but also the latent causes of failures. •Pro-SCHEMe was found to be applicable to Korea NPPs. -- Abstract: The aim of this study is to propose a new quantitative evaluation method for Nuclear Safety Culture (NSC) in Nuclear Power Plant (NPP) operation teams based on the probabilistic approach. Various NSC evaluation methods have been developed, and the Korea NPP utility company has conducted the NSC assessment according to international practice. However, most of methods are conducted by interviews, observations, and the self-assessment. Consequently, the results are often qualitative, subjective, and mainly dependent on evaluator’s judgement, so the assessment results can be interpreted from different perspectives. To resolve limitations of present evaluation methods, the concept of Safety Culture Healthiness was suggested to produce quantitative results and provide faster evaluation process. This paper presents Probabilistic Safety Culture Healthiness Evaluation Method (Pro-SCHEMe) to generate quantitative inputs for Human Reliability Assessment (HRA) in Probabilistic Safety Assessment (PSA). Evaluation items which correspond to a basic event in PSA are derived in the first part of the paper through the literature survey; mostly from nuclear-related organizations such as the International Atomic Energy Agency (IAEA), the United States Nuclear Regulatory Commission (U.S.NRC), and the Institute of Nuclear Power Operations (INPO). Event trees (ETs) and fault trees (FTs) are devised to apply evaluation items to PSA based on the relationships among such items. The Modeling Guidelines are also suggested to classify and calculate NSC characteristics of respective NPPs. Probability of the fault tree top event, namely safety culture healthiness, is automatically calculated to determine the state of NSC healthiness of operation teams. Validation of the suggested method performed by case studies using training video of NPP operators. According to the validation results, a positive relationship between ‘success’ states of safety culture and human performance was found, the safety culture state probability profile of each team represents the team characteristic, and the cut-set analysis of the proposed method provides not only the root causes but also the latent causes of failure. Pro-SCHEMe showed possibility to apply NSC to NPP system safety analysis judging by the results of the case study. Further case studies will be conducted to meet the statistical requirement of the results.
[en] Highlights: •A systematic methodology was used to reflect safety metric under a hybrid of DSA & PSA. •4-type typology (either capacity or load was deterministic or probabilistic) was proposed. •Safety measures of the capacity-load case from CDF’s perspective was illustrated. •A deterministic safety margin with its natural probabilistic counterpart was presented. -- Abstract: Theoretical criteria for design or regulatory safety of nuclear power plants often take the form of requirements that some model of the “capacity” of the plant to respond to a hypothesized threat sufficiently exceed a model of the “load” presumably placed upon the plant by that threat. Either of capacity or load can be deterministic or probabilistic, which leads to a four-type typology, as opposed to the traditional classification of theories of nuclear safety as either deterministic or probabilistic. Concrete examples of each of the four types are provided. Possible uses of this viewpoint for design and regulation are discussed, especially as regards melding of the basically deterministic notion of safety margins with its natural probabilistic counterpart of requiring load exceed capacity with only very small probability. Use of this viewpoint is illustrated by using it as a framework within which to describe the regulatory impact of the well-known ECCS hearings of the 1970s.
[en] There has been a need for upgrading and replacing the current systems of Nuclear Power Plants (NPP) with new technology due to obsolescence and spare parts issues, and demands for higher performance. However, the processes for new technology deployment in NPPs may encounter risks causing unpredictable outcomes leading to performance or operation degradation. Hence, proper Risk Management (RM) is required for ensuring safety and performance of NPPs since it provides a means to identify risks and minimize their impacts. For these reasons, the purpose of this research is to investigate how risks are managed in practice and to propose the proper RM for the deployment of new technology in NPPs.