• Nuclear Engineering and Technology
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• 제51권1호
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• pp.104-115
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• 2019

Since the Fukushima Daiichi accident, there has been an emphasis on the risk resulting from multi-unit accidents. Human reliability analysis (HRA) is one of the important issues in multi-unit probabilistic safety assessment (MUPSA). Hence, there is a need to properly identify all the human and organizational factors relevant to a multi-unit incident scenario in a nuclear power plant (NPP). This study identifies and categorizes the human and organizational factors relevant to a multi-unit incident scenario of NPPs based on a review of relevant literature. These factors are then analyzed to ascertain all possible unit-to-unit interactions that need to be considered in the multi-unit HRA and the pattern of interactions. The human and organizational factors are classified into five categories: organization, work device, task, performance shaping factors, and environmental factors. The identification and classification of these factors will significantly contribute to the development of adequate strategies and guidelines for managing multi-unit accidents. This study is a necessary initial step in developing an effective HRA method for multiple NPP units in a site.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1483-1490
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• 2021

Many studies are undertaken into nuclear power plants (NPPs) in preparation for accidents exceeding design standards. In this paper, we analyze the applicability of various wireless communication technologies as accident countermeasures in different NPP environments. In particular, a commercial wireless communication module (WCM) is investigated by measuring leakage current and packet error rate (PER), which vary depending on the intensity of incident radiation on the module, by testing at a Co-60 gamma-ray irradiation facility. The experimental results show that the WCMs continued to operate after total doses of 940 and 1097 Gy, with PERs of 3.6% and 0.8%, when exposed to irradiation dose rates of 185 and 486 Gy/h, respectively. In short, the lower irradiation dose rate decreased the performance of WCMs more than the higher dose rate. In experiments comparing the two communication protocols of request/response and one-way, the WCMs survived up to 997 and 1177 Gy, with PERs of 2% and 0%, respectively. Since the request/response protocol uses both the transmitter and the receiver, while the one-way protocol uses only the transmitter, then the electronic system on the side of the receiver is more vulnerable to radiation effects. From our experiments, the tested module is expected to be used for design-based accidents (DBAs) of "Category A" type, and has confirmed the possibility of using wireless communication systems in NPPs.

• 한국지진공학회논문집
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• 제19권1호
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• pp.21-28
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• 2015

The nuclear accident due to the recent earthquake in Japan has triggered awareness of the importance of safety with regard to nuclear power plants (NPPs). An earthquake is one of the most important parameters which governs the safety of NPPs among external events. Application of a base isolation system for NPPs can reduce the risk for earthquakes. At present, a soil-structure interaction (SSI) analysis is essential in the seismic design of NPPs in consideration of the ground structure interaction. In the seismic analysis of the base-isolated NPP, it is restrictive to consider the nonlinear properties of seismic isolation devices due to the linear analysis of the SSI analysis programs, such as SASSI. Thus, in this study, SSI analyses are performed using an iterative approach considering the material nonlinearity of the isolators. By performing the SSI analysis using an iterative approach, the nonlinear properties of isolators can be considered. The difference between the SSI analysis results without iteration and SSI with iteration using SASSI is noticeable. The results of the SSI analysis using an effective linear (non-iterative) approach underestimate the spectral acceleration because the effective linear model cannot consider the nonlinear properties of isolators. The results of the SSI analysis show that the horizontal response of the base-isolated NPP is significantly reduced.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Nuclear Engineering and Technology
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• 제34권2호
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• 한국자원식물학회:학술대회논문집
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• 한국자원식물학회 2018년도 추계학술대회
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• pp.25-25
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• 2018

The Fukushima Daiichi Nuclear Power Plant accident in March 2011 caused severe radioactive contamination in the surrounding environment. Since the accident, much attention has been paid to the biological and genetic consequences of organism inhabiting the contaminated area. The effect of radiation exposure on genetic mutation rates is little known, especially for low doses and in situ conditions. Evaluating DNA mutation by low levels of radiation dose is difficult due to the rare mutation event and lack of sequence information before the accident. In this study, correlations with air dose levels and somatic DNA mutation rates were evaluated using Next Generation Sequencer for the clonal plant, Phyllostachys edulis. This bamboo is known to spread an identical clone throughout Japan, and it has the advantage that we can compare genetic mutation rate among identical clone growing different air dose levels. We collected 94 samples of P. edulis from 14 sites with air dose rates from $0.04{\sim}7.80{\mu}Gy/h$. Their clonal identity was confirmed by analysis using 24 microsatellite markers, and then, sequences among samples were compared by MIG sequence. The sequence data were obtained from 2,718 loci. About ~200,000 bp sequence (80 bp X 2,718 loci) were obtained for each sample, and this corresponds to about 0.01% of the genome sequence of P. edulis. In these sequences, 442 loci showed polymorphism patterns including recent origin mutation, old mutation, and sequence errors. The number of mutations per sample ranged from 0 to 13, and did not correlate with air dose levels. This result indicated that DNA mutations have not accumulated in P. edulis living in the air doses levels less than $10{\mu}Gy/h$. Our study also suggests that mutation rates can be assessed by selecting an appropriate experimental approach and analyzing with next generation sequencer.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.541-546
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• 2014

Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

• 에너지공학
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• 제23권3호
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• pp.186-190
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• 2014

원자력발전소는 냉각재상실사고(LOCA)와 같은 과도상태시 pH 조절을 통해 격납건물의 핵분열생성물(요오드) 제거 능력을 유지한다. 이와 더불어 격납건물 내부의 스테인레스강 기기들의 응력부식균열(Stress Corrosion Cracking)을 방지하고 알루미늄 또는 아연 부식에 의한 수소생성을 최소화할 수 있기 때문에 살수 및 집수조냉각수의 화학조건(pH) 조절능력이 요구된다. 현재 원전은 LOCA시 능동형 살수첨가제인 NaOH를 사용하여 격납건물 살수 및 집수조냉각수의 pH를 조절하도록 설계되어있다. 본 논문에서는 LOCA시 집수조냉각수의 pH를 분석하고, 살수화학조건 pH 관련 최신규제요건인 표준심사지침(SRP) 6.5.2에 따라 핵분열생성물제거상수 및 제염계수를 계산하였다. 분석결과, 격납건물집수조 pH는 8.09~9.67로서 설계기준을 만족한다. 그리고 격납건물살수계통에 의한 핵분열생성물 제거상수 및 제염계수는 원전 내환경기기검증을 위한 방사선환경 평가의 입력으로 제공된다.