• 한국기계가공학회지
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• 제18권8호
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• pp.31-37
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• 2019

The inter-development of FMECA is very important to assess the effect of potential failures during system operation on mission, safety and performance. Among these, criticality analysis is a core task that identifies items with high risk and selects the analyzed objects as the key management targets and reflects their effects to the design optimization. In this paper, we analyze the theory related to criticality analysis following US military standard, and propose a method to quantify the failure effect probability for objective criticality analysis. The criticality analysis according to the US military standard depends on the subjective judgment of the failure probability. The methodology for quantifying the failure effect probability is presented by using the reliability theory and the Bayes theorem. The failure rate is calculated by applying the method to quantify failure effect probability.

• Journal of Radiation Protection and Research
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• 제18권2호
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• pp.47-59
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• 1993

본 연구에서는 사용후핵 연료를 안전하게 수송할 수 있는 수송용기의 여러 가지 설계 항목중에 수송용기 내부에 장전한 핵연료에 의한 핵임계반응을 방지하기 위한 핵임계해석을 수행하였다. 핵임계 해석에 사용한 HANSEN-ROACH-KENO-Va 전산시스템에 대한 검증계산을 수행하였고 수송용기의 핵임계측면에서의 안전성을 확보하기 위해 가능한 보수적인 가정을 하여 어떠한 경우에도 수송용기에 장전된 핵연료가 임계상태에 도달하지 않도록 수송용기 내부의 구조 및 적절한 핵임계 방지제를 선택하였고 정상수송 및 가상사고 조건 등에 대한 해석을 수행하였다. 그 결과 KSC-7 수송용기 의 설계조건을 만족하고 핵임계측면에서의 안전성을 보장할 수 있는 재료 및 구조에 대한 결론을 해석적으로 도출하였다.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1148-1155
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• 2020

This work presents two different methods for quantifying and propagating the uncertainty associated with fuel composition at end of life for cask criticality calculations. The first approach, the computational approach uses parametric uncertainty including those associated with nuclear data, fuel geometry, material composition, and plant operation to perform forward depletion on Monte-Carlo sampled inputs. These uncertainties are based on experimental and prior experience in criticality safety. The second approach, the data-driven approach relies on using radiochemcial assay data to derive code bias information. The code bias data is used to perturb the isotopic inventory in the data-driven approach. For both approaches, the uncertainty in k_eff for the cask is propagated by performing forward criticality calculations on sampled inputs using the distributions obtained from each approach. It is found that the data driven approach yielded a higher uncertainty than the computational approach by about 500 pcm. An exploration is also done to see if considering correlation between isotopes at end of life affects k_eff uncertainty, and the results demonstrate an effect of about 100 pcm.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.815-819
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• 2018

In this study, criticality safety assessments of the potential for storing transuranic element (TRU) ingots via a pyroprocess were evaluated to determine the appropriate TRU storage design parameters, in this case the ratio of the TRU ingot height to the radius and the number of TRU ingot canisters stacked within a container. Various accident situations were modeled over a modeling period of 5 years for a cumulative inventory of TRU ingots with various water densities in submerged containers and with various pitches between the containers in the facility. Under these combinations, we calculated the threshold of TRU height and radius ratio depending on the number of canisters in a container to keep the stored TRU in a subcritical state. The ratio of the TRU ingot height to radius should not exceed 4.5, 1.1, 0.5, 0.3, and 0.2 for two, three, four, five, and six levels of stacked canisters in a container, respectively.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.17-26
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• 1986

연소를 고려하는 사용후핵연료저장조의 핵임계 안전성 분석에서 검증용 계산 체제와 rack계산 체제 사이의 편차를 신뢰성 있게 결정하는 방법을 시험하였다. 이를 위하여 고리 1호기의 사용후핵연료저장조를 연소를 고려하는 가장 조밀한 rack으로 개념설계하고, 핵연료의 농축도 및 연소도에 따라 증배계수를 계산하였다. 표준값 생산용 Monte Carlo 코드로는 KENO-IV를 그리고 실제 rack 설계용으로는 2차원 충돌화률 코드인 FATAC을 사용하였다. 이 두 계산의 결과를 상호 비교하여 계산 체제 사이의 편차와 이의 경향성 및 신뢰도를 평가하였다. 이 방법을 사용하면 확실한 신뢰도 근거를 마련할 수 있을 뿐만 아니라 반응도 여유면에서 기존의 방법보다 불리하지 않음이 입증되었다.

• 한국산학기술학회논문지
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• 제16권5호
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• pp.3370-3377
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• 2015

전동차는 대용량 교통수단으로서 정시 운행 및 높은 안전성이 요구되기 때문에, 고장 분석을 체계적으로 수행하여 신뢰도를 향상시키기 위한 수단으로서 고장 영향의 심각도 및 치명도를 정량적으로 평가하는 FMECA (Failure Modes, Effects and Criticality Analysis) 기법이 적용되고 있다. 그러나, 아직까지 전동차에 특화된 FMECA 규격 및 절차는 정립되어 있지 않고 자동차 산업 등 다른 산업분야의 FMECA 규격을 그대로 적용하고 있기 때문에 전동차의 고유한 운영 및 유지보수 여건을 충분히 반영하지 못하고 있다. 본 논문에서는 산업계 각 분야에서 적용되고 있는 FMECA 규격에 대한 분석을 토대로, 전동차 분야에 적합한 FMECA 기법으로서 고장 영향 분석과 치명도 분석을 단계별로 나누어 수행하고 고장 영향의 심각도에 중점을 두어 치명도를 분석하는 기법을 제시하였다. 제안 기법을 전동차의 핵심 안전 장치인 고속차단기에 적용하여 도시철도 현장에서의 15년 동안의 전동차 유지보수 데이터를 이용하여 분석한 결과, 고속차단기 부품 중에서 특히 아크 슈트의 절손이 심각도 3등급, 치명도 5등급으로 위험도가 가장 높았으며, 뒤를 이어서 전자변 파손 및 접촉 불량, 실린더 파손 등이 심각도 3등급, 치명도 4등급으로 위험도가 높은 것으로 나타났다. 이상의 분석 결과는 전동차 고속차단기의 설계 및 유지보수 업무의 개선에 활용할 수 있다.

• International journal of advanced smart convergence
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• 제1권1호
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• pp.57-60
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• 2012

In this paper, we have performed a safety analysis on an automotive evaporator inspection system. We performed the bi-directional analysis on the manufacturing line. Software Fault Tree Analysis (SFTA) as backward analysis and Software Failure Modes, Effects, & Criticality Analysis (SFMECA) as forward analysis are performed alternately to detect potential cause-to-effect relations. The analysis results indicate the possibility of searching and summarizing fault patterns for future reusability.