• Nuclear Engineering and Technology
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• v.50 no.8
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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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• v.53 no.2
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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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• v.44 no.8
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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.

• Journal of Radiation Protection and Research
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• v.36 no.2
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• pp.99-106
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• 2011

Tritium is the one of the most important radionuclide for workers in nuclear power plants (NPPs) and the public, from the dosimetric point of view. Humans are likely to have internal radiation exposure by tritium inhalation. Radiation exposure by tritium accounts for approximately 7% and 60~90% of the total radiation exposure of NPP workers and the public during normal operation, respectively. Thus, many researches have been conducted to estimate the internal dose by tritium precisely in the world. In terms of tritium dosimetry, this paper provides the current status of research for tritium metabolism and dosimetry.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2119-2129
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• 2020

Flow-accelerated corrosion (FAC), liquid droplet impingement erosion (LDIE), cavitation and flashing can cause continuous wall-thinning in nuclear secondary pipes. In order to prevent pipe rupture events resulting from the wall-thinning, most NPPs (nuclear power plants) implement their management programs, which include periodic thickness inspection using UT (ultrasonic test). Meanwhile, it is well known in field experiences that the thickness measurement errors (or deviations) are often comparable with the amount of thickness reduction. Because of these errors, it is difficult to estimate wall-thinning exactly whether the significant thinning has occurred in the inspected components or not. In the previous study, the authors presented an approximate estimation procedure as the first step for thickness measurement deviations at each inspected component and the statistical & quantitative characteristics of the measurement deviations using plant experience data. In this study, statistical significance was quantified for the current methods used for wall-thinning determination. Also, the authors proposed new estimation procedures for determining local wall-thinning to overcome the weakness of the current methods, in which the proposed procedure is based on analysis of variance (ANOVA) method using subgrouping of measured thinning values at all measurement grids. The new procedures were also quantified for their statistical significance. As the results, it is confirmed that the new methods have better estimation confidence than the methods having used until now.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1462-1470
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• 2020

Reliability in safety-critical systems and equipment is of vital importance, so the probabilistic safety assessment (PSA) has been widely used for many years in the nuclear industry to address reliability in a quantitative manner. As many nuclear power plants (NPPs) become digitalized, evaluating the reliability of safety-critical software has become an emerging issue. Due to a lack of available methods, in many conventional PSA models only hardware reliability is addressed with the assumption that software reliability is perfect or very high compared to hardware reliability. This study focused on developing a new method of safety-critical software reliability quantification, derived from hardware-software integrated environment testing. Since the complexity of hardware and software interaction makes the possible number of test cases for exhaustive testing well beyond a practically achievable range, an importance-oriented testing method that assures the most efficient test coverage was developed. Application to the test of an actual NPP reactor protection system demonstrated the applicability of the developed method and provided insight into complex software-based system reliability.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.421-431
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• 2023

Nuclear power plants (NPPs) have technical specifications (Tech Specs) to ensure that the equipment and key operating parameters necessary for the safe operation of the power plant are maintained within limiting conditions for operation (LCO) determined by a safety analysis. The LCO of Tech Specs that identify the lowest functional capability of equipment required for safe operation for a facility must be complied for the safe operation of NPP. There have been previous studies to aid in compliance with LCO relevant to rule-based expert systems; however, there is an obvious limit to expert systems for implementing the rules for many situations related to LCO. Therefore, in this study, we present a retrieval methodology for similar LCO cases in determining whether LCO is met or not met. To reflect the natural language processing of NPP features, a domain dictionary was built, and the optimal term frequency-inverse document frequency variant was selected. The retrieval performance was improved by adding a Boolean retrieval model based on terms related to the LCO in addition to the vector space model. The developed domain dictionary and retrieval methodology are expected to be exceedingly useful in determining whether LCO is met.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.60 no.8
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• pp.1632-1638
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• 2011

Electric cables are one of the most important components in a nuclear power plant since they provide the power needed to operate electrical equipment. Despite their importance, cables typically receive little attention since they are considered passive, long-lived components that have been very reliable over the years when subjected to the environmental conditions for which they were designed. The operating experience reveals that a defect of the insulator or poor construction causes the initial failure of cable. However, the number of cable failures increase with plant aging, and these cable failures are occurring within the plants' 40-year licensed period. These cable failures have resulted in plant transients, shutdown, loss of safety functions or redundancy, entries into limiting conditions for operation, and challenges for plant operators. Therefore, diagnosis of MV cable installed in NPPs has become one of the most urgent issues in recent years. In accordance with PSR, condition maintenance for cables is also continuously required. Recently, HFPD tests have been widely performed to diagnose cable in the transmission and distribution cable system. However, on-line HFPD wasn't used in the NPPs because of the danger of plant shutdown, measurement sensitivity and application problems, etc. In this paper, HFPD measurement with portable device was performed to evaluate the integrity of the 4.16kV & 13.8kV cable lines. The test results show that HFPD is highly attractive to the diagnosis of MV cables in NPP by high detection sensitivity on-site.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.110-116
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• 2022

Human failure event (HFE) dependency analysis is a part of human reliability analysis (HRA). For efficient HFE dependency analysis, a maximum number of minimal cut sets (MCSs) that have HFE combinations are generated from the fault trees for the probabilistic safety assessment (PSA) of nuclear power plants (NPPs). After collecting potential HFE combinations, dependency levels of subsequent HFEs on the preceding HFEs in each MCS are analyzed and assigned as conditional probabilities. Then, HFE recovery is performed to reflect these conditional probabilities in MCSs by modifying MCSs. Inappropriate HFE dependency analysis and HFE recovery might lead to an inaccurate core damage frequency (CDF). Using the above process, HFE recovery is performed on MCSs that are generated with a non-zero truncation limit, where many MCSs that have HFE combinations are truncated. As a result, the resultant CDF might be underestimated. In this paper, a new method is suggested to incorporate HFE recovery into the MCS generation stage. Compared to the current approach with a separate HFE recovery after MCS generation, this new method can (1) reduce the total time and burden for MCS generation and HFE recovery, (2) prevent the truncation of MCSs that have dependent HFEs, and (3) avoid CDF underestimation. This new method is a simple but very effective means of performing MCS generation and HFE recovery simultaneously and improving CDF accuracy. The effectiveness and strength of the new method are clearly demonstrated and discussed with fault trees and HFE combinations that have joint probabilities.

• Journal of Welding and Joining
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• v.26 no.2
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• pp.43-49
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• 2008

Fatigue cracks due to thermal stratification or corrosion in pipelines of nuclear power plants can cause serious problems on reactor cooling system. Therefore, the development of an integrated technology including fabrication of standard specimens and their practical usage is needed to enhance the reliability of nondestructive testing. The test material was austenitic STS 304, which is used as pipelines in the Reactor Coolant System of a nuclear power plants. The best condition for fabrication of thermal fatigue cracks at the notch plate was selected using the thermal stress analysis of ANSYS. The specimen was installed from the tensile tester and underwent continuos tension loads of 51,000N. Then, after the specimen was heated to $450^{\circ}C$ for 1 minute using HF induction heater, it was cooled to $20^{\circ}C$ in 1 minute using a mixture of dry ice and water. The initial crack was generated at 17,000 cycles, 560 hours later (1cycle/2min.) and the depth of the thermal fatigue crack reached about 40% of the thickness of the specimen at 22,000 cycles. As a results of optical microscope and SEM analysis, it is confirmed that fabricated thermal fatigue cracks have the same characteristics as real fatigue cracks in nuclear power plants. The crack shape and size were identified.