• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.53-54
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• 2019

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.99-100
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.100-101
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.35-36
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• 2018

• Nuclear Engineering and Technology
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• 제36권5호
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• pp.420-429
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• 2004

The local power density should be estimated accurately to prevent fuel rod melting. The local power density at the hottest part of a hot fuel rod, which is described by the power peaking factor, is more important information than the local power density at any other position in a reactor core. Therefore, in this work, the power peaking factor, which is defined as the highest local power density to the average power density in a reactor core, is estimated by fuzzy neural networks using numerous measured signals of the reactor coolant system. The fuzzy neural networks are trained using a training data set and are verified with another test data set. They are then applied to the first fuel cycle of Yonggwang nuclear power plant unit 3. The estimation accuracy of the power peaking factor is 0.45% based on the relative $2_{\sigma}$ error by using the fuzzy neural networks without the in-core neutron flux sensors signals input. A value of 0.23% is obtained with the in-core neutron flux sensors signals, which is sufficiently accurate for use in local power density monitoring.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2454-2463
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• 2021

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2006년도 정기 학술대회 논문집
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• pp.828-835
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• 2006

Spacer grid springs support the fuel rods in a nuclear fuel system. The spacer grid is a part of a fuel assembly. Since a spring has repeated contacts with the fuel rod, fretting wear occurs on the surface of the spring. Design is usually performed to reduce the wear. The conceptual design process for the spring is defined by using the Independence of axiomatic design and the design is carried out based on the direction that the design matrix indicates. For detailed design an optimization problem is formulated. In optimization, homologous design is employed to reduce fretting wear. The deformation of a structure is called homologous if a given geometrical relationship holds for a given number of structural points before, during, and after the deformation. In this case, the deformed shape of the spring should be the same as that of the fuel rod. 1bis condition is transformed to a function and considered as a constraint in the optimization process. The objective function is minimizing the maximum stress to allow a local plastic deformation. Optimization results show that the contact occurs in a wide range. Also, the results are verified by nonlinear finite element analysis.

• 분석과학
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• 제12권2호
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• pp.94-98
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• 1999

Quadrupole Mass Spectrometer를 이용하여 핵연료봉으로부터 포집된 1기업이하 소량의 기체들로부터 그들의 조성과 동위원소비를 구하는 방법을 검토하였다. He, $N_2$, $O_2$, Ar, Kr, Xe의 개별기체와 혼합기체를 이용하여 기체압력과 조성비에 따른 검정곡선의 직선성을 조사하였다. Sample chamber와 analyser chamber 사이에 부착된 molecular leak의 영향을 조사하였으며, 시료와 유사한 조성을 갖는 혼합표준기체로부터 각 기체의 감도를 얻은 후 동일조건에서 시료를 분석하였다. 측정압력 범위에서 Kr과 Xe의 동위원소간 감도차는 크게 나타나지 않았다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.27-36
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• 2020

Natural convection and thermal stratification phenomena are found in large water pools that are being used as heat sinks for decay heat removal from the reactor core using passive heat removal systems. In this study, the two-phase (water and air) natural convection and thermal stratification phenomena with conjugate heat transfer in the rectangular enclosure were investigated numerically using ANSYS Fluent 17.2 code. The transient numerical simulations of these phenomena in the full-scale computational domain of the experimental facility were performed. Generation of water vapour bubbles around the heater rod and evaporation phenomena were included in this numerical investigation. The results of numerical simulations are in good agreement with experimental measurements. This shows that the natural convection is formed in region above the heater rod and the water is thermally stratified in the region below the heater rod. The heat from higher region and from the heater rod is transferred to the lower region via conduction. The thermal stratification disappears and the water becomes well mixed, only after the water temperature reaches the saturation temperature and boiling starts. The developed modelling approach and obtained results provide guidelines for numerical investigations of thermal-hydraulic processes in the water pools for passive residual heat removal systems or spent nuclear fuel pools considering the concreate walls of the pool and main room above the pool.