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Monte Carlo 방법을 이용한 바나듐 자발 중성자계측기 초기 민감도 계산
차균호 ( Kyoon Ho Cha ),박영우 ( Young Woo Park ) 한국센서학회 2016 센서학회지 Vol.25 No.3
Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Topmount In-core Instrumentation) which might be used in the SMART and SMR.
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이서정,차균호,Lee, Seo Jeong,Cha, Kyoon Ho 한국방사성폐기물학회 2018 방사성폐기물학회지 Vol.16 No.3
A subcritical facility must ensure nuclear criticality safety under all circumstances. For this purpose, it is essential to have a procedure to validate that calculated values do not exceed upper subcritical limit (USL), determined by quantifying the bias and uncertainty. However, there are several validation methodologies of nuclear criticality safety and these can yield different USL. Therefore, it is necessary to analyze the validity of the methodologies to establish one methodology that can provide the most appropriate USL. In this study, two documents, a guide for validation of nuclear criticality safety calculational methodology (NUREG/CR-6698) and a criticality benchmark guide for light water reactor fuel in transport and storage package (NUREG/CR-6361), are compared and analyzed. In particular, the methodology in NUREG/CR-6361 is applied to the USLSTATS code. However, the analysis results show that the methodology in NUREG/CR-6698 is more appropriate, for several reasons. This is applied to decision of USL to design casks using SCALE code version 6.1.
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사용후핵연료 안전관리를 위한 정보 검증 및 분석 모듈 구축
유승욱(Seung Uk Yoo),심가희(Ga-Hee Sim),은정무(Jeong Mu Eun),차균호(Kyoon-Ho Cha) 한국에너지학회 2024 에너지공학 Vol.33 No.1
원전의 사용후핵연료는 계획된 처분절차에 따라 원전 내 습식/건식 저장시설부터 중간저장시설을 거쳐 영구처분시설까지 인도, 인수기관 및 규제기관 등 다양한 기관들이 관여한다. 만약 사용후핵연료의 잘못된 정보가 타 기관에 전달되면 처분까지의 과정에서 다양하고 심각한 문제가 발생할 수 있기 때문에 사용후핵연료 안전정보의 검증은 필수적이다. 본 논문은 XGBoost 기반의 이상치 검증을 통해 핵연료 데이터에 대해 인적오류 여부를 확인한 후 ORIGEN 코드를 통해 이상치 검증 절차에서 확인된 핵연료 데이터를 2차 검증하여, 마지막으로 2차 검증된 핵연료 데이터를 기반으로 사용후핵연료 운반/저장 용기에 대한 방사선학적 안전성을 평가하는 등 세 가지 방법으로 사용후핵연료 안전정보의 무결점성을 검증하는데 있다. According to the planned disposal procedure, spent nuclear fuel passes through various institutions from temporary storage facilities to permanent disposal facilities. Conveying incorrect information regarding spent fuel to other institutions can result in serious issues throughout the disposal process. Therefore, validation of safety information regarding spent nuclear fuel is one of the necessary issues in the disposal process. This paper employs three method to validate the data. First of all, it confirms the presence of human errors in nuclear fuel data through anomaly detection based on XGBoost. Then, it conducts a secondary validation of the nuclear fuel data identified during the anomaly detection process using the ORIGEN code. Finally, based on the double- validated nuclear fuel data, validation module analyzes the radiological safety of the transportation and storage casks for spent nuclear fuel.
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