• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.88-96
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• 2018

Concerns over reliability assessments of the main components in nuclear power plants (NPPs) related to aging and continuous operation have increased. The conventional reliability assessment for main components uses experimental correlations under general conditions. Most NPPs have been operating in Korea for a long time, and it is predictable that NPPs operating for the same number of years would show varying extent of aging and degradation. The conventional reliability assessment does not adequately reflect the characteristics of an individual plant. Therefore, the reliability of individual components and an individual plant was estimated according to operating data and conditions. It is essential to reflect aging as a characteristic of individual NPPs, and this is performed through prognostics. To handle this difficulty, in this paper, the general path model/Bayes, a data-based prognostic method, was used to update the reliability estimated from the generic database. As a case study, the authors consider the aging for steam generator tubes in NPPs and demonstrate the suggested methodology with data obtained from the probabilistic algorithm for the steam generator tube assessment program.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.105-132
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• 2011

The concept of improved performance indicators (PIs) for use in the KINS Safety Performance Indicator (SPI) program for reactor safety area is proposed in this paper. To achieve this, the recently developed PIs from the USNRC that use risk information were investigated, and a feasibility study for the application of these PIs in Korean NPPs was performed. The investigated PIs are Baseline Risk Index for Initiating Events (BRIIE), Unplanned Scrams with Complications (USwC), and Mitigating System Performance Index (MSPI). Moreover, the thresholds of the existing safety performance indicators of KINS were evaluated in consideration of the risk and regulatory response to different levels of licensee performance in the graded inspection program.

• Journal of the Korean Society of Safety
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• v.29 no.6
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• pp.172-178
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• 2014

In terms of an accident management, the cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results since the Three Mile Island (TMI) accident. The objectives of this paper are to explain how to identify the plant response and cope with its vulnerabilities using the probabilistic safety assessment (PSA) quantified results and severe accident database SARDB(Severe Accident Risk Data Bank) based on sequences analysis results. Although PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behaviour. The plant model used in this paper is oriented to the cases of loss of coolant accident (LOCA) is be used as a training simulator for a severe accident management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Structural Engineering and Mechanics
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• v.72 no.1
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• pp.31-41
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• 2019

Structural integrity assessment of piping components is of paramount important for remaining life prediction, residual strength evaluation and for in-service inspection planning. For accurate prediction of these, a reliable fracture parameter is essential. One of the fracture parameters is stress intensity factor (SIF), which is generally preferred for high strength materials, can be evaluated by using linear elastic fracture mechanics principles. To employ available analytical and numerical procedures for fracture analysis of piping components, it takes considerable amount of time and effort. In view of this, an alternative approach to analytical and finite element analysis, a model based on relevance vector machine (RVM) is developed to predict SIF of part through crack of a piping component under fatigue loading. RVM is based on probabilistic approach and regression and it is established based on Bayesian formulation of a linear model with an appropriate prior that results in a sparse representation. Model for SIF prediction is developed by using MATLAB software wherein 70% of the data has been used for the development of RVM model and rest of the data is used for validation. The predicted SIF is found to be in good agreement with the corresponding analytical solution, and can be used for damage tolerant analysis of structural components.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3298-3313
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• 2021

The objective of this study is to estimate the occurrence probabilities of severe accident phenomena based on a fuzzy elicitation technique. Normally, it is difficult to determine these probabilities due to the lack of information on severe accident progression and the highly uncertain values currently in use. In this case, fuzzy set theory (FST) can be best exploited. First, questions were devised for expert elicitation on technical issues of severe accident phenomena. To deal with ambiguities and the imprecision of previously developed (reference) probabilities, fuzzy aggregation methods based on FST were employed to derive the occurrence probabilities of severe accidents via four phases: 1) choosing experts, 2) quantifying weighting factors for the experts, 3) aggregating the experts' opinions, and 4) defuzzifying the fuzzy numbers. In this way, this study obtained expert elicitation results in the form of updated occurrence probabilities of severe accident phenomena in the OPR-1000 plant, after which the differences between the reference probabilities and the newly acquired probabilities using fuzzy aggregation were compared, with the advantages of the fuzzy technique over other approaches explained. Lastly, the impact of applying the updated severe accident probabilities on containment integrity was quantitatively investigated in a Level 2 PSA model.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1394-1405
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• 2022

The importance of ensuring the inherent safety and security has been more emphasized in recent years to demonstrate the integrity of nuclear facilities under external human-induced events (e.g. aircraft crashes). This work suggests a simulation methodology to effectively evaluate the impact of a commercial aircraft engine onto a dry storage facility. A full-scale engine model was developed and verified by Riera force-time history analysis. A reinforced concrete (RC) structure of a dry storage facility was also developed and material behavior of concrete was incorporated using three constitutive models namely: Continuous Surface Cap, Winfrith, and Karagozian & Case for comparison. Strain-based erosion limits for concrete were suitably defined and the local responses were then compared and analyzed with empirical formulas according to variations in impact velocity. The proposed methodology reasonably predicted such local damage modes of RC structure from the engine missile, and the analysis results agreed well with the calculations of empirical formulas. This research is expected to be helpful in reviewing the dry storage facility design and in the probabilistic risk assessment considering diverse impact scenarios.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.229-232
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• 2003

A waste container, one of the key compartments in a multi-barrier system for a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic strata and the swelling pressure of the bentonite buffer. Also, it prohibits potential release of radionuclides for a certain period of time. before it is corroded by impurities. Even though the materials of a waste container is carefully chosen and all manufacturing processes are under heavy quality assurance, there might be a slight chance of intial defects in a waste container. Also, during the deposition of a waste container in a repository, there might be a chance of an incident affecting the integrity of a waste container. In this study, the FEP's and the scenarios over radiological impact of a potential initial waste container defect was developed. Then the total system performance assessment on this initial waste container failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set studied in this paper, the annual individual dose by the ICF scenario well meets the KINS regulation.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.244-251
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• 2015

The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1&2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0~35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2~3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.