• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.525-525
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• 2005

This study is for the seismic analysis and the structural integrity evaluation of the box beam for supporting nuclear fuel-transfer-basket of the HANARO reactor pool. For performing the seismic analysis and evaluating the structural integrity in air or submerged condition, the finite element model of the fuel-transfer-basket and its supporting box beam(the coupled model) was developed. The hydrodynamic effect is also considered by using added mass concept. The seismic response spectrum analyses of the coupled model under the design floor response spectrum loads of Safe Shutdown Earthquake(SSE) were performed. Through the numerical experiments, the analysis results show that the stress values of the coupled model lot the structural integrity are within the ASME Code limits.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.564-569
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• 2003

During fuel irradiation tests, all parts of cylindrical structure with multiple holes act as heat sources due to fussion heal and ${\gamma}-flux$. The high temperature is especially generated over $2500^{\circ}C$ in the center of pellet. Due to the high temperature, many problems occur, such as melting of pellet and declining of heat transfer between cladding and coolant. [n this study, it is attempted 10 minimize the temperature of pellet using optimization method about geometric variables. For thermal and optimization analysis or structure. the finite element method code. ANSYS 5.7 is used. In this procedure. subproblem approximation method is used to the optimization methods. Through the optimum design process, the temperature of sealed basket type is reduced from $2537^{\circ}C$ to $2181^{\circ}C$ and the temperature of open basket type is reduced from $2560^{\circ}C$ to $2106^{\circ}C$.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.763-764
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• 2004

• 방사성폐기물학회지
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• 제21권3호
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• pp.359-370
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• 2023

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

• Journal of Radiation Protection and Research
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• 제28권4호
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• pp.281-290
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• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.166-177
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• 1993

사용후 핵연료 건식 저장 캐니스터의 핵연료 바스켓 안에 있는 CANDU 37소자 핵연료 다발의 최대 온도를 계산하기 위한 단순화된 열해석 방법과 함께 대기 조건 인자들이 캐니스터 내부의 최대 핵연료 온도에 미치는 영향을 조사하기 위해 수행한 표본 해석 결과를 제시하였다. 3가지 모우드(mode)의 열전달이 공존하는 캐니스터 내부핵연료 다발의 복잡한 기하학적 구조에 대한 다차원열전달 문제를 풀기 위하여 건식저장 캐니스터에 저장된 CANDU 사용후 핵연료 다발들을 등가 및 동심의 핵연료 실린더(cylinder)로 대치하였다. 추가적인 입력 자료 및 열전달 상관식을 이용하여 등가 핵연료 실린더의 단순화된 축대칭, 2차원, 복수 모우드(multi-mode)의 열전달 문제를 기존의 컴퓨터 코드인 HEATING5로 해석하였다. 예측한 온도 분포와 식물 모형 실험 결과의 비교는 만족스러울 정도로 일치함을 보여주고 있다.

• 대한기계학회논문집A
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• 제28권10호
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• pp.1509-1516
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• 2004

During fuel irradiation tests, all parts of cylindrical structure with multiple holes act as heat sources due to fussion heat and ${\gamma}$-flux. The high temperature is especially generated in the center of pellet. Because of the high temperature, many problems occur, such as melting of pellet and declining of heat transfer between cladding and coolant. In this paper, it is attempted to minimize the temperature of pellet using optimization method. For thermal and optimization analysis of structure, the finite element method code, ANSYS 5.7 is used. Through the optimum design process, the temperature of SBT diminished 10％ and the temperature of OBT diminished 18％.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.131-144
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• 2010

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology for recycling useful resources from spent fuel since 1997. The process includes pretreatment, electroreduction, electrorefining, electrowinning, and a waste salt treatment system. This paper briefly addresses unit processes and related innovative technologies. As for the electroreduction step, a stainless steel mesh basket was applied for adaption of granules of uranium oxide. This basket was designed for ready handling and transfer of feed material. A graphite cathode was used for the continuous collection of uranium dendrite in the electrorefining system. This enhances the throughput of the electrorefiner. A particular mesh type stirrer was designed to inhibit uranium spill-over at the liquid Cd crucible. A residual actinide recovery system was also tested to recover TRU tracer. In order to reduce the waste volume, a crystallization method is employed for Cs and Sr removal. Experiments on the unit processes were tested successfully, and based on the results, engineering-scale equipment has been designed for the PRIDE (PyRoprocess Integrated inactive DEmonstration facility).

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.969-978
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• 2013

A full-sized model for the horizontally oriented metal cask containing 21 spent fuel assemblies has been considered to evaluate the internal natural convection behavior within a dry shield canister (DSC) filled with helium as a working fluid. A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values ($3{\times}10^6{\sim}3{\times}10^7$) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results. The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to $1.5{\times}10^6$ < Ra < $1.0{\times}10^7$ are proportional to 0.5 power of the Rayleigh number, while the Nusselt numbers for $1.0{\times}10^7$ < Ra < $8.0{\times}10^7$ are proportional to 0.27 power of the Rayleigh number. These results agreed well with the trends of the experimental data for Ra > $1.0{\times}10^7$.