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.

Transport packages have been developed to transport the decommissioning waste from the nuclear power plant. The packages are classified with Type IP-2 package. The IAEA requirements for Type IP-2 packages include that a free drop test should be performed for normal conditions of transport. In this study, drop tests of the packages were performed to prove the structural integrity and to verify the reliability of the analysis results by comparing the test and analysis results. Half-scale models were used for the drop tests and drop position was considered as 0.3 m oblique drop on packages weighing more than 15 tons. The strain and impact acceleration data were obtained to verify the reliability of the analysis results. Before and after the drop tests, radiation shielding tests were performed to confirm that the dose rate increase was within 20% at the external surface of the package. Also, measurement of bolt torque, and visual inspection were performed to confirm the loss or dispersion of the radioactive contents. After each drop test, slight deformations occurred in some packages. However, there was no loss of pretension in the lid bolts and the shielding thickness was not reduced for metal shields. In the package with concrete shield, the surface dose rate did not increase and there was no cracks or damage to the concrete. Therefore, the transport packages met the legal requirements (no more than a 20% increase of radiation level and no loss or dispersion of radioactive contents). Safety verifications were performed using the measured strain and acceleration data from the test, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, it was found that the structural integrity of the packages was maintained under the drop test conditions. The results of this study were used as design data of the transport packages, and the packages will be used in the NPP decommissioning project in the future.

Spent nuclear fuels released from the reactor are stored in cooling pools and then stored in dry storage casks. During the transition from the wet storage to dry storage cask, a vacuum drying process is used to remove residual water in the cask. During the vacuum drying process, gas pressure is reduced to below 400 Pa to promote evaporation and water removal. KAERI is developing a PWR single assembly (PLUS7) test equipment to simulate the thermal flow in spent fuel assembly. In this study, the thermal conductivity of air at low pressure was derived to perform the thermal analysis of the canister in vacuum. In addition, thermal analyses were performed for the canister with backfill gases of helium, air, and a vacuum in the vertical orientation using the COBRA-SFS code. At low pressure, the thermal conductivity of air depends on pressure and temperature. The reduced thermal conductivity, kr (W/m-K) was calculated using the curve fit for air at reduced pressure in thin gaps presented in the General Electric Fluid Flow Handbook. 􀝇􀯥/􀝇􀬴 = 􀬵 􀬵􀬾 􀮼􀯍/􀯉􀰋 Where, k0 is the thermal conductivity at atmospheric pressure (W/m-K), P is the reduced (vacuum) pressure (Pa), δ is the gap size (m), T is the temperature (K), and C is the Lasance constant (7.657E-5 N/m-K). The thermal conductivity of air decreases as the pressure decreases. The reduced thermal conductivity of air at pressures of 400 Pa and 40 Pa was calculated to be 0.97 and 0.77, respectively. For the analysis in vacuum, no enhancement of the convective heat transfer was assumed (Nu=1.0). For the helium backfill, the peak cladding temperature was the lowest and the axial temperature profile was the flattest due to the higher thermal conductivity and lower density of the helium. For the vacuum backfill, the peak cladding temperature was the highest and temperature gradient was the sharpest due to the only radiative heat transfer effect in the fuel assembly.

Waste containers for packaging, transportation and disposal of NPP (Nuclear Power Plant) decommissioning wastes are being developed. In this study, drop tests were conducted to prove the safety of containers for packaging of the wastes and to verify the reliability of the analysis results by comparing the test and analysis results. The drop height of the waste containers was considered to be 30 mm, which is the maximum lifting speed of a 50 tons crane in the waste treatment facility converted to the drop height. Drop orientation of the containers was considered for bottom-end on drop. The impact acceleration and strain data were obtained to verify the reliability of the analysis results. Before and after the drop tests, measurement of the dose rate and the radiographic testing for concrete wall, and measurement of the wall thickness of steel plate were conducted to evaluate the radiation shielding integrity. Also, measurement of bolt torque, and visual inspection were conducted to evaluate the loss or dispersion of radioactive contents. After the drop tests, the radiation dose rate on the container surface did not increase by more than 20%, and there was no crack in the concrete. In addition, the thickness of the steel plate did not change within the measurement error. Therefore, the radiation shielding integrity of the container was maintained. After the drop tests, the lid bolts were not damaged and there was no loss of pretension in the lid bolts. In addition, there was no loss or dispersion of the contents as a result of visual inspection. In order to prove the reliability of the drop analysis results, safety verifications were performed using the drop test results, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, the structural integrity of the waste containers was maintained under the drop test conditions.

Detailed temperature distributions of the spent fuel are required to evaluate the long-term integrity of the dry storage system. In this study, a subchannel analysis method was established to obtain the detailed temperatures of a spent fuel using the COBRA-SFS code. The SAHTT (Single Assembly Heat Transfer Test) model was selected as the subchannel analysis. It was developed at the PNL to investigate heat transfer characteristics of spent PWR fuel under dry storage conditions. The SAHTT has a 15×15 rod array with simulated rods 0.42 in. (10.7 mm) in diameter. Control rod thimbles were modeled with unheated rods. The COBRA-SFS input consists a detailed subchannel model with 256 subchannels, 225 rods, and 8 slab nodes. The heat generation rate was axially uniform with total power of 1.0 kW. Subchannel analyses were performed for the vertical orientation under three different backfills of air, helium, and vacuum. For the vacuum backfill, the peak temperature was the highest and temperature gradients the sharpest only due to the radiation heat transfer effect. For the helium backfill, peak temperature was lowest and the axial profiles flattest due to the higher conductivity and lower density of helium. Subchannel analyses were also performed to evaluate the effect of thermal parameters such as surface emissivity, convective heat transfer coefficients, and flow resistance coefficients on the PCT (Peak Cladding Temperature). The PCT was affected by the emissivity of the fuel rod and the basket, and in particular, the basket emissivity had a greater effect. The PCT was affected by the Nusselt number, but the range of the Nusselt number is around 3.66. Therefore, the effect of the Nusselt number on the PCT will not be significant. As a result of the analysis according to the flow resistance coefficients, the PCT was affected by the wall friction factor, but the loss coefficients from the space grid had little effect. Subchannel technique obtained from this work can be used to predict the detailed temperature distributions of spent fuel assembly.

Bird screen meshes are installed at the air inlet and outlet ducts of spent fuel storage casks to inhibit the intrusion of debris from the external environment. The presence of these screens introduces an additional resistance to air flow through the ducts. In this study, a porous media model was developed to simplify the bird screen meshes. CFD analyses were used to derive and verify the flow resistance factors for the porous media model. Thermal analyses were carried out for concrete storage cask using the porous media model. Thermal tests were performed for concrete casks with bird screen meshes. The measured temperatures were compared with the analysis results for the porous model. The analysis results agreed well with the test results. The analysis temperatures were slightly higher than the test temperatures. Therefore, the reliability and conservatism of the analysis results for the porous model have been verified.

IAEA 및 국내의 방사성물질 운반 관련 규정에 따라 중·저준위 방사성폐기물 드럼 8개를 운반할 수 있는 IP-2형 운반용기를 개발하였다. IP-2형 운반용기는 낙하시험 및 적층시험을 거친 후 내용물의 유실 또 는 분산과 운반용기 외부표면에서의 방사선량률이 20 % 이상 증가할 수 있는 차폐능력의 상실이 없어야 한다. 본 연구의 목적은 적층시험조건에 대한 시험방법 및 절차를 수립하고 IP-2형 운반용기의 적층조건 에 대한 구조적 건전성을 평가하는데 있다. 운반용기의 원형시험모델을 이용하여 운반용기 중량의 5배 하중으로 24시간 동안 압축하는 적층조건에 대한 시험 및 전산해석을 수행하였다. 적층시험 시 운반용기 의 모서리기둥에서의 변형률 및 변위를 측정하였으며, 측정된 변형률 및 변위는 해석결과와 서로 일치하 였다. 컨테이너 바닥부의 처짐량은 측정이 어렵기 때문에 전산해석 방법으로 구하였다. 모서리기둥의 최 대 변위와 컨테이너 바닥의 최대 처짐은 법규에서 규정하는 허용치에 비하여 낮게 나타났다. 적층시험 전?후에는 운반용기의 외형치수, 차폐체 두께, 볼트토크 등을 측정하였으며, 그 값들을 비교분석한 결과 운반용기는 내용물의 유실 및 분산, 차폐체 두께의 감소가 나타나지 않았다. 따라서 적층시험조건에서 IP-2형 운반용기의 구조적 건전성이 입증되었다.

사용후연료 건식저장용기의 전복사고조건에 대한 1/3 축소모델의 시험을 실시하여 전복해석에 대한 검증을 하였다. 전복해석은 전복각도에 따른 위치에너지와 동일한 운동에너지를 가지는 초기각속도를 이용하여 결정된 각 점에서의 속도를 충돌직전 모델에 대한 초기경계값으로 입력하여 해석하였다. 전복시험에 따른 캐니스터의 구조적 건전성을 확인하기 위하여 육안검사와 함께 액체침투법과 초음파 탐상법와 같은 비파괴검사를 실시하였다. 전복충격에 의하여 저장용기의 뚜껑 에 변형 이 발생되었지만 캐니스터의 구조적 건전성이 유지되었다. 시험에서 취득한 변형률과 가속도를 해석결과와 비교하여 해석 에 대한 검증을 실시하였다. 해석결과는 시험결과보다 대체로 두 배 정도의 큰 값을 주는 것으로 나타났다.

사용후핵연료 건식저장용기는 낙하사고조건에서 캐니스터의 건전성이 입증되어야 한다. 낙하사고조건은 캐니스터를 건식저장용기에 장입하기 위하여 저장용기의 상부에서 크레인으로 취급하는 도중에 캐니스터가 저장용기 내부의 받침대로 자유 낙하하는 조건이다. 저장용기 내부의 받침대는 이러한 조건에서 캐니스터의 구조적 건전성을 유지하도록 완충효과가 좋아야 한다. 본 연구에서는 다양한 저장용기 내부 받침대 에 대한 3차원 유한요소해석을 통하여 낙하사고조건에서 캐니스터의 구조적 건전성을 향상시킬 수 있는 구조를 결정하였다. 저장용기 내부 받침대는 탄소강으로 만들어진 원통 쉘의 내부에 콘크리트를 장입한 구조와 받침대 높이의 변화 없이 콘크리트 높이의 1/4정도에 탄소강과 폴리우레탄폼을 이용한 구조물을 사용하여 완충효과를 보완하고자 수정된 구조를 고려하였다. 완충체의 형상 및 구조를 결정하기 위하여 십자형상이나 원형의 탄소강 구조물을 받침대 상부에 위치하여 그 영향을 알아보았다. 이때 탄소강 구조물의 두께를 24 mm, 12 mm, 6mm로 변화를 주었다. 또한, 탄소강 구조물 사이에 충진하는 폴리우레탄폼의 밀도에 대한 영향을 알아보았다.

This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 under the normal condition. The off-normal condition has an environmental temperature of 38 . An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.