• Nuclear Engineering and Technology
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• 제38권6호
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• pp.489-504
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• 2006

A review has been made for the previous studies on safety of a geologic repository for high-level radioactive wastes (HLW) related to autocatalytic criticality phenomena with positive reactivity feedback. Neutronic studies on geometric and materials configuration consisting of rock, water and thermally fissile materials and the radionuclide migration and accumulation studies were performed previously for the Yucca Mountain Repository and a hypothetical water-saturated repository for vitrified HLW. In either case, it was concluded that it would be highly unlikely for an autocatalytic criticality event to happen at a geologic repository. Remaining scenarios can be avoided by careful selection of a repository site, engineered-barrier design and conditioning of solidified HLW. Thus, criticality safety should be properly addressed in regulations and site selection criteria. The models developed for radiological safety assessment to obtain conservatively overestimated exposure dose rates to the public may not be used directly for the criticality safety assessment, where accumulated fissile materials mass needs to be conservatively overestimated. The models for criticality safety also require more careful treatment of geometry and heterogeneity in transport paths because a minimum critical mass is sensitive to geometry of fissile materials accumulation.

• 한국안전학회지
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• 제31권5호
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• pp.7-15
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• 2016

The purpose of this paper is to carry out Failure Modes and Effects Analysis (FMEA) and use criticality in order to determine risk priority number of the components of electric power installations in Engineering college building of D university. In risk priority number, GROUP A had 7 failure modes; more specifically, Transfomer had 4 modes, Filter(C)(1 mode), LA(1 mode), and CB(MCCB)(1 mode), and thus 4 components had failure modes. In terms of criticality, high-grade group a total of 16 failure modes, and 7 components-LA(1 mode), CB(MCCB)(1 mode), MOF(2 modes), PT(1 mode), Transformer(7 modes), Cable(3 modes), and Filter(C)(1 mode)-had failure modes. Comparison of risk priority number and criticality was made. The components which had high risk priority number and high criticality were Transformer, Filter(C), LA, and CB(MCCB). The components which had high criticality were MOF and cable. In particular, Transformer(RPN: 4 modes, Criticality: 7 modes) was chosen as an intensive management component.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• 정보처리학회논문지:소프트웨어 및 데이터공학
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• 제10권8호
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• pp.287-300
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• 2021

협업 사이버-물리 시스템(Collaborative Cyber-Physical Systems, CCPS)은 물리 세계와 사이버 세계가 밀접하게 결합하여 공동의 목표를 달성하기 위하여 협업을 수행하는 시스템이다. 한편, 단일 사이버-물리 시스템(Cyber-Physical System)의 경우에는 ISO 26262 또는 IEC 61508과 같은 표준을 따르거나 다양한 위험 분석 기법을 적용함으로써 그 안전을 확보할 수 있다. 그러나 CCPS에서는 협업을 수행중인 한 CPS의 결함으로 인하여 다른 협업 중인 CPS에게 수많은 결함을 발생시키기 때문에 안전의 확보가 매우 어렵다. 본 논문에서는 이러한 CCPS의 위험을 분석하여 안전을 확보하기 위해 복합적인 위험 분석과 위험 분석 산출물 사이의 관계를 기반으로 하는 위험 치명도 매트릭스(Fault Criticality Matrix, FCM)를 제시한다. FCM에서는 결함, 결함의 치명도, 안전 가드와 안전 가드의 발생 확률, 결함의 영향 및 순위를 나열하여 분석한다. 안전 엔지니어는 이를 통해 시스템의 설계 단계에서 각 결함의 치명도와 영향을 분석하고, 설계된 안전 가드를 통해 식별된 고장을 효과적으로 관리하고 제어함으로써 안전한 CPS를 개발할 수 있다. 제시된 방법의 유용성을 확인하기 위해 CCPS의 대표적 예인 군집주행에 대하여 사례 연구를 수행하였다. 본 연구에서 개발된 도구를 사용하여 군집주행 시스템에 FCM을 적용함으로써 상세한 결함 치명도 분석을 수행하였고, 분석 결과는 적합성과 효과성 관점에서 점검되었다. 또한 군집 주행에 대한 시뮬레이션 수행을 통해 FCM을 사용하여 결함 치명도를 분석한 군집주행 시스템이 발견된 모든 결함을 완화시켜 충돌 가능성을 크게 낮추었음을 보였다.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.19-29
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• 2021

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

• 방사성폐기물학회지
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• 제16권3호
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• pp.315-330
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• 2018

미임계 시설은 정상 또는 사고상태에서 핵임계안전성이 확보되어야 한다. 이를 위해선 계산된 임계도가 바이어스와 불확실도로 결정된 미임계상한치(USL)를 초과하지 않는다는 것을 검증하는 절차가 반드시 필요하다. 하지만 핵임계안전성 검증방법론은 여러 가지가 존재하며, 방법론이 달라지면 USL도 달라지므로 가장 적절한 한가지의 방법론으로 평가하는 것이 중요하다. 본 연구에서는 핵임계안전성 검증 방법론이 기술된 두 개의 문서를 비교 분석하여 한 가지 방법론으로 정립하였고, SCALE6.1 코드를 이용한 용기 설계에서의 미임계상한치 결정에 적용하였다.

• 한국컴퓨터정보학회:학술대회논문집
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• 한국컴퓨터정보학회 2019년도 제60차 하계학술대회논문집 27권2호
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• pp.7-8
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• 2019

In this paper, we applied the migration technique to real-time tasks that have relatively low criticality but still important to be dropped by the mixed-criticality scheduling algorithms. The proposed drop and migrate algorithm analyzes the schedulability by calculating CPU utilization and response time of using task migration. We provide analysis to guarantee the deadline of LO-tasks, by transforming the response time equation specified with migration time. The transformed response time equation was able to analyze the migration schedulability. This algorithm can be used with various mixed-criticality schedulers as a supplementary method. We expect this algorithm will be used for scheduling LO-tasks such as communication task that requires safety guarantee especially in platooning and autonomous driving by utilizing the advantages of multiple node connectivities.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• 방사성폐기물학회지
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• 제13권2호
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• pp.141-154
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• 2015

경수로 사용후핵연료 수송/저장용기의 핵임계 해석은 사용후핵연료내의 악티나이드핵종 및 핵분열생성물 함유량에 대한 불확실성을 이유로 신연료로 가정된 가상의 연료를 선정하여 평가해오고 있다. 그러나 이러한 평가방법은 용기 설계 시 과도한 임계여유도를 유도하여 경제적 손실을 유발할 수 있는 단점이있다. 이와 같은 단점을 극복하기 위하여 최근 연소도이득효과(burnup credit, BUC)를 반영한 수송저장용기의 설계 및 상용화를 위한 연구가 추진되었다. 이에 본 연구에서는 한국원자력환경공단에서 개발중인 금속겸용용기를 대상으로 연소도 이득효과적용 시 핵임계 안전성(criticality safety)에 영향을 미칠 것으로 예상되는 '노심 운전인자', '축방향 연소도 분포', '오장전 사고상황'에 대하여 핵임계 평가를 수행하였다. 그 결과 노심운전인자 중 저농축, 고연소도일 때 비출력에 따른 핵임계 변화가 크게 평가되었으며, 고연소도 사용후핵연료에서 End effect가 양의 값을 나타내었다. 특히 오장전에 의한 유효증배계수는 최대 0.18467증가하였으므로, 연소도이득효과를 적용 할 경우 필수고려사항임을 확인하였다. 본 연구결과는 국내모델(금속겸용용기)의 연소도 이득효과 적용기술 개발 및 사용 후핵연료 장전 시 일어날 수 있는 오장전 사고를 방지하기 위한 운영절차 개발에 참고자료로 활용될 수 있다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1708-1714
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• 2019

The purpose of the critical evaluation of the spent fuel pool (SFP) is to verify that the maximum effective multiplication factor ($K_{eff}$) is less than the critical safety limit at 100% stored condition of the spent fuel with the maximum reactivity. At nuclear power plants, the storage standard of spent fuel, ie, the loading curve, is established to prevent criticality from being generated in SFP. Here, the loading curve refers to a graph showing the minimum discharged burnup versus the initial enrichment of spent fuel. Recently, US NRC proposed the new critical safety assessment guideline (DSS-ISG-2010-01, Revision 0) of PWR SFPs and most of utilities in US is following it. Of course, the licensed criterion of the maximum effective multiplication factor of SFP remains unchanged and it should be less than 0.95 from the 95% probability and the 95% confidence level. However, the new guideline is including the new evaluation methodologies like the application of the axial burnup profile, the validation of depletion and criticality code, and trend analysis. Among the new evaluation methodologies, the most important factor that affects $K_{eff}$ is the axial burnup profile of spent fuel. US NRC recommends to consider the axial burnup profiles presented in NUREG-6801 in criticality analysis. In this paper, criticality effect was evaluated considering three profiles, respectively: i) Axial burnup profiles presented in NUREG-6801. ii) Representative PWR axial burnup profile. iii) Uniform axial burnup profile. As the result, the case applying the axial burnup profiles presented in NUREG-6801 showed the highest $K_{eff}$ among three cases. Therefore, we need to introduce a new methodology because it can be issued if the axial burnup profiles presented in NUREG/CR-6801 are applied to the domestic nuclear power plants without any other consideration.