• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.1702-1707
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• 2007

The reduction of pipe-thickness induced by flow accelerated corrosion (FAC) is one of the most serious problems on the maintenance of piping system in nuclear power plants (NNP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain pressure and consequently results in leakage or rupture. For this reason, wall thinning by FAC has been inspected in secondary side piping systems in NPPs. In this research Round Robin Test (RRT) was conducted to verify confidence of wall thinning measurement system in NPP. 12 inspectors from 3 companies participated and 23 specimens were used according to standard practice in RRT. The gage R&R analysis was introduced in regard to repeatability and reproducibility that are affected to measurement system errors. Confidence intervals of thickness measurement system were obtained.

• Journal of Magnetics
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• 제16권2호
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• pp.88-91
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• 2011

Steam generator tubes (SGTs) in nuclear power plants (NPPs) are a boundary between the primary side generating heat by nuclear fission and the secondary side generating electric power by a turbine. The water inside the SGT is high temperature and high pressure. Therefore, defects and magnetic phases (MPs) are partly produced in non-magnetic SGT by high stresses and temperatures. This causes trouble regarding the safety of SGTs but it is difficult to detect the MP using the conventional eddy current technique (ECT). In particular, a circumferential defect (CD) and circumferential magnetic phase (CMP) cannot detected by ECT. Consequently, a new method is needed to detect CDs and CMPs in SGT. A new U-type yoke with two types of coils was designed and the reactance signal by the CMPs and CDs in the SGT material was simulated.

• 대한인간공학회지
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• 제34권1호
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• pp.45-62
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• 2015

Objective: The purpose of this study is to investigate the effect of personality type on human performance tool compliance in nuclear power plants (NPPs) and to propose general recommendations for an enhancement of its practical utilization. Background: Various guidelines, regulating criteria, and recommendations have been developed to prevent human errors in NPPs. Despite these efforts, the accidents sometimes caused by human errors have steadily occurred, and therefore, various human performance tools have been adopted as countermeasures against human errors. The major and inevitable contributing factors among the many hazards to human errors might be the trait and personality, which are considered to be the inner world of humans. Thus, we try to investigate the utilization of human performance tools by considering the different types of operating crew personalities, and we suggested more practical recommendations to prevent human errors according to the personality. Method: We developed the Questionnaire using the Big 6 (HEXACO) models, which are human performance tools for workers in NPPs, and individual (condition) variables to investigate the effect of personality types on human performance tools. We slightly modified them to help the survey respondents understand them better. A survey was conducted for ordinary people over the age of 20. SPSS 22.0 was used to perform a correlation analysis and a hierarchical regression analysis to find the relationship between personality types and human performance tools. Results: The utilization of human performance tools shows significant differences statistically by personality. The correlation result reveals that the types of Honesty (H), Extraversion (X), Conscientiousness (C), and Openness to experience (O) show a higher utilization of human performance tools. In hierarchical regression results, human performance tools of task preview, questioning attitude, stopping when unsure, self-checking, effective communication, and place-keeping show a higher utilization with personality types. However, the Agreeableness (A) type did not show significant differences statistically with human performance tools. Conclusion: We tried to investigate the utilization of human performance tools by considering the different types of human personality and provide more practical recommendations to prevent human errors according to the personality. These results will be able to prevent human errors owing to the characteristics (advantages and disadvantages) of personality types. Application: This information can be utilized as guidelines for proactive recommendations according to the workers' personalities for more practical human performance tools to prevent human errors in an NPP.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.764-775
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• 2020

Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.

• 한국안전학회지
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• 제35권1호
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• pp.87-96
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• 2020

The purpose of this paper is to develop a fire HRA (Human Reliability Analysis) procedure for full power operation of domestic NPPs (Nuclear Power Plants). For the development of fire HRA procedure, the recent research results of NUREG-1921 in an effort to meet the requirements of the ASME/ANS PRA Standard were reviewed. The K-HRA method, a standard method for HRA of a domestic level 1 PSA (Probabilistic Safety Assessment) and fire related procedures in domestic NPPs were reviewed. Based on the review, a procedure for the fire HRA required for a domestic fire PSA based on the K-HRA method was developed. To this end, HRA issues such as new operator actions required in the event of a fire and complexity of fire situations were considered. Based on the four kinds of HFE (Human Failure Event) developed for a fire HRA in this research, a qualitative analysis such as feasibility evaluation was suggested. And also a quantitative analysis process which consists of screening analysis and detailed analysis was proposed. For the qualitative analysis, a screening analysis by NUREG-1921 was used. In this research, the screening criteria for the screening analysis was modified to reduce vague description and to reflect recent experimental results. For a detailed analysis, the K-HRA method and scoping analysis by NUREG-1921 were adopted. To apply K-HRA to fire HRA for quantification, efforts to modify PSFs (Performance Shaping Factors) of K-HRA to reflect fire situation and effects were made. For example, an absence of STA (Shift Technical Advisor) to command a fire brigade at a fire area is considered and the absence time should be reflected for a HEP (Human Error Probability) quantification. Based on the fire HRA procedure developed in this paper, a case study for HEP quantification such as a screening analysis and detailed analysis with the modified K-HRA was performed. It is expected that the HRA procedure suggested in this paper will be utilized for fire PSA for domestic NPPs as it is the first attempt to establish an HRA process considering fire effects.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1184-1191
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• 2016

This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs), because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA). This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.709-718
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• 2019

The international nuclear industry has undergone a lot of changes since the Fukushima, Chernobyl and TMI nuclear power plant accidents. However, there are still large and small component deficiencies at nuclear power plants in the world. There are many causes of electrical equipment defects. There are also factors that cause component failures due to human errors. This paper analyzed the root causes of failure and types of human error in 300 cases of electrical component failures. We analyzed the operating experience of electrical components by methods of root causes in K-HPES (Korean-version of Human Performance Enhancement System) and by methods of human error types in HuRAM+ (Human error-Related event root cause Analysis Method Plus). As a result of analysis, the most electrical component failures appeared as circuit breakers and emergency generators. The major causes of failure showed deterioration and contact failure of electrical components by human error of operations management. The causes of direct failure were due to aged components. Types of human error affecting the causes of electrical equipment failure are as follows. The human error type group I showed that errors of commission (EOC) were 97%, the human error type group II showed that slip/lapse errors were 74%, and the human error type group III showed that latent errors were 95%. This paper is meaningful in that we have approached the causes of electrical equipment failures from a comprehensive human error perspective and found a countermeasure against the root cause. This study will help human performance enhancement in nuclear power plants. However, this paper has done a lot of research on improving human performance in the maintenance field rather than in the design and construction stages. In the future, continuous research on types of human error and prevention measures in the design and construction sector will be required.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1289-1297
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• 2018

Preliminary surface soil Derived Concentration Guideline Levels (DCGLs) were derived conforming to the Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM) procedure for the site release and reuse of Kori Unit 1 in Korea. Based on the decommissioning experiences of the U.S. nuclear power plants, a suite of residual radionuclides was determined, and uncertainties contributed to the resultant dose by the input parameters were quantified via the sensitivity analysis of parameters. The peak of the mean dose was obtained via the probabilistic analysis of the RESRAD (RESidual RADioactivity)-ONSITE code. Consequently, $DCGL_w$ of Kori Unit 1 in accordance with two scenarios, industrial worker and residential farmer scenario, were derived and the results were compared respectively with other NPPs. It could be used as a basic guideline for establishing regulatory standards for reuse planning, designing the site characterization surveys and implementing final status survey (FSS).

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.357-367
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• 2021

In general, common cause failures (CCFs) have been modeled with the assumption that components within the same group are symmetric. This assumption reduces the number of parameters required for the CCF probability estimation and allows us to use a parametric model, such as the alpha factor model. Although there are various asymmetric conditions in nuclear power plants (NPPs) to be addressed, the traditional CCF models are limited to symmetric conditions. Therefore, this paper proposes the copulabased CCF model to deal with asymmetric as well as symmetric CCFs. Once a joint distribution between the components is constructed using copulas, the proposed model is able to provide the probability of common cause basic events (CCBEs) by formulating a system of equations without symmetry assumptions. In addition, Bayesian inferences for the parameters of the marginal and copula distributions are introduced and Markov Chain Monte Carlo (MCMC) algorithms are employed to sample from the posterior distribution. Three example cases using simulated data, including asymmetry conditions in total failure probabilities and/or dependencies, are illustrated. Consequently, the copula-based CCF model provides appropriate estimates of CCFs for asymmetric conditions. This paper also discusses the limitations and notes on the proposed method.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.897-910
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• 2012

Inappropriate communications can cause a lack of necessary information exchange between operators and lead to serious consequences in large process systems such as nuclear power plants (NPPs). In this regard, various kinds of taxonomies of inappropriate communications have been developed to prevent inappropriate communications. However, there seems to be difficult to identify inappropriate communications from verbal protocol data between operators. Because the existing taxonomies were developed for use in report analysis, there is a problem of 'uncertainty'. In consequence, this paper proposes a new taxonomy of inappropriate communications and provides some insights to prevent inappropriate communications. In order to develop the taxonomy, existing taxonomies for four industries from 1980 to 2010 were collected and a new taxonomy is developed based on the simplified one-way communication model. In addition, the ratio of inappropriate communications from 8 samples of audio-visual format verbal protocol data recorded during emergency training sessions by operating teams is compared with performance scores calculated based on the task analysis. As a result, inappropriate communications can be easily identified from the verbal protocol data using the suggested taxonomy, and teams with a higher ratio of inappropriate communications tend to have a lower performance score.