• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.139-152
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• 2006

A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.22 no.4
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• pp.826-833
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• 1998

The energy-absorbing characteristic of impact limiters affects the cask design so significantly that it should be evaluated as accurate as possible. The objective of this study is to find the influence of the impact limiter's steel case and gusset plates which enclose the shock absorbing cellular material on the impact energy absorption. The influence of impact limiter's steel case and gusset plate stiffeners on the impact energy absorption behavior under horizontal drop impact was evaluated for a radioactive isotope transport cask. Though the impact limiters mitigate the impact damage of the cask, the impact limiter's steel case and gusset plate stiffeners increase the impact force so significantly that should be designed as soft as possible. The impact analysis without considering impact limiter's steel case and gusset plates stiffener gives non-conservative results, so the stiffness of the steel case and gusset plates should be considered in impact analysis.

• Journal of the Korea Furniture Society
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• v.19 no.6
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• pp.420-427
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• 2008

To study the suitability of chestnut as a cask for ripening traditional liquors, the absorption property for pure water, liquor with 30% alcohol content and ethanol with 95% alcohol content were investigated for five species. Oriental oak had the greatest amount of liquid absorption while chestnut had the smallest amount. The absorption amount linearly increased with increasing soaking time. The absorption amounts for each section were greatest in RT plane, and the difference between LR plane and LT plane was very small for all soaking conditions. The anisotropy of absorption amount for five species was greatest in Japanese cedar and was smallest in white mulberry on the whole. And the change of absorption amounts according to soaking conditions tended to decrease in softwoods and increase in hardwoods, and the difference among wood species was not clear. From this result, it was found that chestnut with a small absorption amount, regarding of soaking behavior, was a good material as a cask for ripening traditional liquors, whereas small diameter oriental oak with a great absorption amount was not suitable in this purpose.

• Nuclear Engineering and Technology
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• v.55 no.7
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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.

• Journal of computational fluids engineering
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• v.21 no.4
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• pp.33-39
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• 2016

Generally, thermal analysis of spent fuel storage cask has been conducted using the porous media and effective thermal conductivity model to simplify the structural complexity of spent fuel assemblies. As the fuel assembly is composed of two regions; active fuel region corresponding to UO2 pellets and unactive fuel region corresponding to the top and bottom nozzle, the heat transfer performance can be influenced depending on porous media application at these regions. In this study, numerical analysis on concrete storage cask of spent fuel was performed to investigate heat transfer effects for two cases; one was porous media application only to active fuel region(case 1) and the other one was porous media to whole length of fuel assembly(case 2). Using computational fluid dynamics code, the three dimensional, 1/4 symmetry model was constructed. For two cases, maximum temperatures for each component were evaluated below the allowable limits. For the case 1, maximum temperatures for fuel cladding, neutron absorber and baskets inside the canister were slightly higher than those for the case 2. In particular, even though the helium flows with low velocity due to buoyant forces occurred at the top and bottom of unactive fuel region, treating only active fuel region as the porous media was ineffective in respect of the heat removal performance of concrete storage cask, implying a conservative result.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.245-253
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• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.56-65
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• 1989

The nuclear criticality of the KSC-4 shipping cask which can load four assemblies of PWR spent fuel was analyzed using KENO-IV computer code and 19-group nuclear cross section set generated from 218-group neutron cross section library(DLC-43/CSRL) using AMPX system. In accordance with 10CFR71, the analysis was performed for fresh fuel assemblies, instead of the spent fuels, under both norml transportation and hypothetical accident conditions. The calculated maximum multiplication factors(Keff) of the KSC-4 cask were 0.85289 and 0.94185 for the normal transportation and hypothetical accident conditions, respectively. The highest Keff of the KSC-4 cask is within the subcritical limit prescribed in l0CFR71 and accordingly the KSC-4 cask is safely designed in terms of nulear criticality.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.3
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• pp.245-254
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• 2005

Radioactive material is used in the various fields. The numbers of transport for radioactive material have been gradually increased in both domestic and International regions. The safety of the cask should be secured to safely transport of radioactive material. The korean atomic law and the IAEA safety standards prescribe regulations lot the safe transport of radioactive material The cask for spent fuel is comprised of the body and the impact limiter. In this study, the empirical equation of the cask impact force is proposed based on the dimensional analysis. Using this empirical equation the characteristics of the impact limiter are analyzed. The results are also validated by comparing with the previous results of the impact area method and the finite element analysis. The present method can be used to predict the impact force of the cask.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.27-35
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• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.