• Journal of Radiation Protection and Research
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• 제42권2호
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• pp.130-140
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• 2017

Background: With the need for a domestic level 3 probabilistic safety assessment (PSA), it is essential to develop a Korea-specific code. Health effect assessments study radiation-induced impacts; in particular, long-term health effects are evaluated in terms of cancer risk. The objective of this study was to analyze the latest cancer risk models developed by foreign organizations and to compare the methodology of how they were developed. This paper also provides suggestions regarding the development of Korean cancer risk models. Materials and Methods: A review of cancer risk models was carried out targeting the latest models: the NUREG model (1993), the BEIR VII model (2006), the UNSCEAR model (2006), the ICRP 103 model (2007), and the U.S. EPA model (2011). The methodology of how each model was developed is explained, and the cancer sites, dose and dose rate effectiveness factor (DDREF) and mathematical models are also described in the sections presenting differences among the models. Results and Discussion: The NUREG model was developed by assuming that the risk was proportional to the risk coefficient and dose, while the BEIR VII, UNSCEAR, ICRP, and U.S. EPA models were derived from epidemiological data, principally from Japanese atomic bomb survivors. The risk coefficient does not consider individual characteristics, as the values were calculated in terms of population-averaged cancer risk per unit dose. However, the models derived by epidemiological data are a function of sex, exposure age, and attained age of the exposed individual. Moreover, the methodologies can be used to apply the latest epidemiological data. Therefore, methodologies using epidemiological data should be considered first for developing a Korean cancer risk model, and the cancer sites and DDREF should also be determined based on Korea-specific studies.

• 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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• 제53권2호
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• pp.509-521
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• 2021

In this paper, from a review of the size distribution of the bubbles during pool scrubbing obtained from experiments by EPRI, we apply the bubble size distributions to analyses on the decontamination factors of pool scrubbing via I-COSTA (In-Containment Source Term Analysis). We perform sensitivity studies of the bubble size on the various mechanisms of deposition of aerosol particles in pool scrubbing. We also perform sensitivity studies on the size distributions of the bubbles depending on the diameters at the nozzle exit, the molecular weights of non-condensable gases in the carrier gases, and the steam fractions of the carrier gases. We then perform analyses of LACE-ESPANA experiments and compare the numerical ~ results to those from SPARC-90 and experimental results in order to show the effect of the bubble size distributions.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.644-654
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• 2024

After the Tohoku earthquake and tsunami (Japan, 2011), regulatory efforts to mitigate external hazards have increased both the safety requirements and the total capital cost of nuclear power plants (NPPs). In these circumstances, identifying not only disaster robustness but also cost-effective capacity setting of NPPs has become one of the most important tasks for the nuclear power industry. A few studies have been performed to relocate the seismic capacity of NPPs, yet the effects of multiple hazards have not been accounted for in NPP capacity optimization. The major challenges in extending this problem to the multihazard dimension are (1) the high computational costs for both multihazard risk quantification and system-level optimization and (2) the lack of capital cost databases of NPPs. To resolve these issues, this paper proposes an effective method that identifies the optimal multihazard capacity of NPPs using a multi-objective genetic algorithm and the two-stage direct quantification of fault trees using Monte Carlo simulation method, called the two-stage DQFM. Also, a capacity-based indirect capital cost measure is proposed. Such a proposed method enables NPP to achieve safety and cost-effectiveness against multi-hazard simultaneously within the computationally efficient platform. The proposed multihazard capacity optimization framework is demonstrated and tested with an earthquake-tsunami example.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.154-164
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• 2003

The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through Level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities.

• 한국안전학회지
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• 제37권4호
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• 한국식품과학회지
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• 제37권3호
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• pp.484-491
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• 2005

정량적 미생물 위해평가(Quantitative microbial risk assessment: QMRA)는 국민건강에 영향을 주는 잠재된 위해를 연구하여 식품내 존재하는 병원성미생물과 관련한 위해를 체계적으로 평가하는 것이다. 본 연구는 깁밥에서의 Staphylococcus aureus에 대한 QMRA 모델을 개발하고 이를 식품위생관리에서 이용할 수 있는 기준을 제시하여, 식품안전 분야에서의 QMRA의 필요성과 활용성을 알리기 위해서 실시하였다. QMRA 모델은 매장에서부터 최종소비에 이르기까지 4 단계로 구성되었으며, 미생물 성장모델과 조사자료 그리고 확률분포가 김밥의 최종소비에서의 S. aureus 수준을 평가하기 위하여 이용되었다. S. aureus에 대한 양-반응모델이 없는 관계로 최종 소비단계에서의 S. aureus의 오염수준을 잠재적인 위해를 결정하는데 이용하였다. 이를 위하여 5 log CFU/g이상을 잠재적 유해수준으로 가정하였으며, 시뮬레이션 결과 최종 소비되는 김밥에서 이 유해수준을 초과할 가능성은 30.7%로 나타났다. 김밥에서의 S. aureus의 오염수준은 평균 2.67 log CFU/g으로 나타났으며, 민감도 분석에서는 매장에서의 김밥 보관온도 및 시간이 가장 중요한 요인으로 결정되었다. 이러한 결과를 종합하여 볼 때 김밥 매장에서는 현실적으로 보존시간 관리가 어렵다고 한다면 보관온도를 $10^{\circ}C$ 이하로 유지하는 것이 가장 중요한 것으로 나타났다. 본 연구에서와 같이 QMRA는 식품 내 존재할 수 있는 잠재적인 위해에 영향을 미치는 인자들에 대한 평가에 이용될 수 있으며 이를 식품위생관리에 직접적으로 활용 가능한 것으로 나타났다.삼의 분석방법별 기준인 ginsenoside -Rg1과 -Re의 함량비($Rg1/Re{\Leq}3.87$)에 부합되었다.도에서 MA 저해 효과는 쑥갓>미나리>참깨의 순으로, 각각 54, 48, 29%를 나타냈다. 참깨는 20, $100{\mu}g/mL$의 농도에서처럼 가장 작은 효과를 보여줬고, 쑥갓은 50% 이상의 항산화 효과를 나타냈다. Aldehyde/Carboxylic acid assay에서는 참깨가 가장 높은 효과를 보여줬지만 Lipid MA asaay에서는 그에 비해 가장 낮은 효과를 나타냈다.안전한 수준인 것으로 판단된다. 보여진다.ificantly more inclusive. As a result of the evolution of new fibers, materials, processes and markets, we assert that a new "ENGINEERING WITH FIBERS" (EwF)(중략)web.cnu.ac.kr/~fabric이다. 제작된 멀티미디어를 실제 수업에 활용한 결과 수강생(32명)의 96.9%가 보조자료로 사용된 멀티미디어 콘텐츠자료가 실험관련 교과목 수업에 효과적이라고 응답하였고, 87.5%가 활용된 멀티미디어 콘텐츠 자료에 만족하며, 75%가 기존의 교과서와 비교하여 더 많이 활용하였다고 응답하였다. 따라서 멀티미디어 콘텐츠를 활용한 교육은 개인차에 따른 개별화 학습을 가능하게 할 뿐만 아니라 능동적인 참여를 유도하여 학습효율을 높일 수 있을 것으로 기대된다.향은 패션마케팅의 정의와 적용범위를 축소시킬 수 있는 위험을 내재한 것으로 보여진다. 그런가 하면, 많이 다루어진 주제라 할지라도 개념이나 용어가 통일되지

• 한국식품과학회지
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• 제44권1호
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• pp.106-113
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• 2012

국내 식품의 수은 함량 실태를 검토하여 우리나라 국민의 중금속 노출수준에 따른 안전성을 평가하고자 하였다. 우리나라 국민 전체의 중금속 노출수준을 파악하기 위하여 Monte-Carlo simulation에 기반을 둔 확률론적(probabilistic) 위해평가를 실시하였다. 노출평가를 통해 추정된 인구집단의 식이를 통한 중금속 노출량으로부터 JECFA에서 제시한 PTWI 대비 위해도(%)를 산출하여 노출수준의 위해정도를 확인하였다. 본 연구에서 중금속 안전성 평가를 위한 대상 식품 선정은 식품의약품안전청에서 2000년대에 수행한 중금속 관련 연구과제 중 중금속 함량 원시자료가 확보된 178 식품 품목, 17,965건에 대하여 실시하였다. 식품섭취량 및 체중은 질병관리본부에서 발간한 '국민건강영양조사 4기 2차년도(2008년)' 자료를 활용하였다. 수은 함량은 농산물이 0.115(과실류)-45.448(버섯류) ${\mu}g/kg$이었고 육류는 3.723 ${\mu}g/kg$, 수산물은 9.344(극피 척색류)-194.914(어류) ${\mu}g/kg$, 가공식품에는 0.680(주류)-4.412(가공식품) ${\mu}g/kg$이었다. 식품을 통한 수은 섭취량은 4.29 ${\mu}g/kg$으로 PTWI 대비 13.6% 수준이었으며, 극단(P95) 섭취량은 12.48 ${\mu}g/day$로 PTWI 대비 39.7% 수준이었다. 따라서, 우리나라 국민의 식이를 통한 수은 노출은 위해우려가 낮은 수준이었으며 이는 제외국과 유사하거나 낮은 수준이었다.

• 한국수자원학회논문집
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• 제47권10호
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• pp.853-865
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• 2014

수문학적 댐 위험도 분석은 복잡한 수문분석과 연계되어 있으며, 기본적으로 수문분석 과정과 모형에 사용되는 입력자료에 대한 불확실성을 평가하는 과정이 필요하다. 그러나 체계적인 불확실성 분석 과정을 통한 댐 위험도 분석 절차에 대한 연구는 상대적으로 적은편이다. 이러한 점에서 본 연구에서는 기존 연구에 대해서 2가지 주요 개선점을 도출하여 댐 위험도 분석에 활용하였다. 첫째, 강우 분석시 매개변수의 불확실성 분석이 가능한 Bayesian 모형 기반의 지역빈도해석 절차를 수립하였다. 둘째, 강우-유출 모형 매개변수의 사후분포를 정량적으로 추정하기 위하여 Bayesian 모형과 연계한 HEC-1모형을도입하였다. 도출된 유입 시나리오를 댐의 수위로 환산하기 위하여 기존 저수지 운영기준에 근거하여 저수지 추적을 수행하였으며, 최종적으로 실행함수를 통하여 수문학적 위험도를 추정하였다. 실제 댐에 대해서 모형의 적합성을 평가하였으며, 초기수위 가정에 따른 수문학적 위험도에 민감도를 평가하였다.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2534-2546
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• 2021

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.