• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2604-2612
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• 2023

Spent nuclear fuel and long-lived radioactive waste must be carefully handled before disposing them off to a geological repository. After the pre-storage period in water pools, spent nuclear fuel is stored in casks, which are widely used for interim storage. Interim storage in casks is very important part in the whole cycle of nuclear energy generation. This paper presents the results of the numerical study that was performed to evaluate the thermal behavior of a metal-concrete CONSTOR RBMK-1500 cask loaded with spent nuclear fuel and placed in an open type interim storage facility which is under fire conditions (steady-state, fire, post-fire). The modelling was performed using the ANSYS Fluent code. Also, a local sensitivity analysis of thermal parameters on temperature variation was performed. The analysis demonstrated that the maximum increase in the fuel load temperatures is about 10 ℃ and 8 ℃ for 30 min 800 ℃ and 60 min 600 ℃ fires respectively. Therefore, during the fire and the post-fire periods, the fuel load temperatures did not exceed the 300 ℃ limiting temperature set for an RBMK SNF cladding for long-term storage. This ensures that fire accident does not cause overheating of fuel rods in a cask.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.175-180
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• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.265-270
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• 1996

The temperature distribution in the pellet was obtained from beginning the general heat conduction equation. The thermal conductivity of pellet used the SIMFUEL data that made clear the effect of burnup on the thermal conductivity degradation. Since the pellet rim acts as the thermal barrier to heat flow. the pellet was subdivided into several rings in which the outer ring was adjusted to play almost the same role as the rim. The local burup in each ring except the outer ring was calculated from the power depression factor based on FASER results. whereas the rim burnup at the outer ring was achieved by the pellet averaged burnup based on the empirical relation. The rim changed to the equivalent Xe film so the predicted temperature shooed the thermal jump across the rim. The observed temperature profiles depended on linear heat generation rate. fuel burnup. and power depression factor. The thermal conductivity degradation modelling can be applied to the fuel performance code to high burnup fuel,

• Korean Chemical Engineering Research
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• 제61권2호
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• pp.328-339
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• 2023

Methanol steam reforming (MSR) is a promising method for hydrogen supplying as a critical step in hydrogen fuel cell commercialization in mobile applications. Modelling and understanding of the reactor behavior is an attractive research field to develop an efficient reformer. Three-layer feed-forward artificial neural network (ANN) and Box-Behnken design (BBD) were used to modelling of MSR process using the Cu-SiO₂ aerogel catalyst. Furthermore, impacts of the basic operational variables and their mutual interactions were studied. The results showed that the most affecting parameters were the reaction temperature (56%) and its quadratic term (20.5%). In addition, it was also found that the interaction between temperature and Steam/Methanol ratio is important on the MSR performance. These models precisely predict MSR performance and have great agreement with experimental results. However, on the basis of statistical criteria the ANN technique showed the greater modelling ability as compared with statistical BBD approach.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.358-373
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• 1995

중수로용 개량핵 연료집합체인 CANFLEX 핵연료다발의 CANDU-6 원자로 장전시 열수송계통에 대한 유동안정성이 분석되었다. CANFLEX 핵연료다발은 기존의 37개봉 핵연료다발과 원자로출력 및 압력강하 측면에서 거의 일치되며, 이로인해 수력적 거동이 양립하는 반면, CANFLEX핵연료다발은 기존의 37개봉 핵연료다발 보다 임계채널 출력이 증가하며, 반경방향 출력분포의 평탄화로 인해 균일한 엔탈피 분포를 확보할 수 있게 된다. CANFLEX 핵연료다발 및 출구모관들의 상호연결관에 대한 SOPHT 모델을 개발하였으며, 이 모델을 이용하여 CANFLEX 핵연료다발이 장전된 월성 1호기의 유동 안정성 거동이 해석되었다. 해석결과, 열수송계통의 출구모관들의 상호연결관이 없을 경우에는 기존의 37개봉 핵연료다발과 같이 유동이 불안정함을 보였으며, 출구모관들의 상호연결관이 있을 경우에는 정격출력의 $\pm$1% 내에서 안정함을 보였다. 따라서 CANFLEX 핵연료다발의 월성 1호기 장전시 열수송계통의 유동안정성 측면에서는 건전할 것으로 판단되었다.

• 방사성폐기물학회지
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• 제5권4호
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• pp.323-338
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• 2007

방사성폐기물 심지층 처분 시스템의 성능과 안정성을 평가하기 위해서는 처분장 환경에서의 열적, 역학적, 수리적, 화학적 거동에 대한 이해와 함께 이들 상호간의 영향을 파악하여야 한다. 복잡한 수학 모델과 모델링 기법을 요하는 THMC 복합거동에 대한 해석을 보다 효과적으로 수행하기 위해 DECOVALEX 국제공동연구가 진행되고 있다. 1992년 이후 4단계에 걸친 국제공동연구를 통해 다양한 조건에서의 THMC 복합거동을 해석하는 기법이 개발되어 왔다. 본 연구에서는 DECOVALEX의 주요 내용 및 현황을 정리하고 향후 참여방안 및 참여효과에 대해 논의한다.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• 에너지공학
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• 제6권1호
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• pp.87-95
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• 1997

금속염화물과 암모니아간의 가역반응을 이용한 화학열펌프는 비프레온계 냉동 냉장시스템으로서 환경규약의 제한이 없고 가스, 전기 및 산업폐열 등 다양한 구동열원을 사용할수 있으며 축열에 의한 에너지 저장이나 산업공정에서의 대용량 에너지관리 시스템 등 응용분야가 다양한 장점을 갖고 있다. 통상 소규모의 실험실 장치에서 파일롯트 플랜트(pilot plant)로 전환하는 과정에서 시스템의 성능에 대한 해석이 필요하며 화학반응기의 거동에 대한 컴퓨터 모사도 필수적이다. 따라서 본 연구에서는 작동조건에 따라 화학식 냉동기 성능계수가 어떻게 변하는지를 예측하였고 반응기 모델링에 의한 동적모사를 수행하여 반응기의 온도거동, 작동조건에 따른 전화율 변화, 반응혼합물의 제조변수가 냉동기 성능에 미치는 영향 등을 고찰하였다.