• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.148-154
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• 1985

Radiation shield design for a shipping cask, KSC-1, was evaluated to verify that the cask can be used in the transportation of a spent fuel assembly discharged from KNU 5 & 6. Radiation source term of the spent fuel assembly was calculated with the computer program ORIGEN-79, QAD-CG, ANISN-KA and DOT 3.5 codes Were used in the shielding calculations and the nuclear cross section data needed was extracted from the DLC-23/CASK library. It is concluded that KSC-1 shipping cask satisfies the requirements specified in the relevant regulations under normal conditions of transport and under accident conditions in transport.

• 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 Radiation Protection and Research
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• v.21 no.4
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• pp.313-327
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• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.32 no.2
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• pp.75-86
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• 2019

The biggest obstacle in the nuclear power generation is the high level radioactive waste such as the spent nuclear fuel. High level radioactivities and generated heat make the safe treatment of the spent nuclear fuel very difficult. Nowadays, the only treatment method is a deep geological disposal technology. This paper treats the structural safe design problem of the spent nuclear fuel disposal canister which is one of the core technologies of the deep geological disposal technology. Especially, this paper executed the nonlinear structural analysis for the stresses and deformations occurring in the canister due to the impulsive force applied to the spent nuclear fuel disposal canister in the case of an accidental drop and ground impact event from the transportation vehicle in the repository. The main content of the analysis is about that the impulsive force is obtained using the commercial rigid body dynamic analysis computer code, RecurDyn, and the stress and deformation caused by this impulsive force are obtained using the commercial finite element static structural analysis computer code, NISA. The analysis results show that large stresses and deformations may occur in the canister, especially in the rid or the bottom of the canister, due to the impulsive force occurring during the collision impact period.

• Korean Chemical Engineering Research
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• v.50 no.4
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• pp.696-701
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• 2012

Non-aqueous processes have been developed for stable management and reuse of spent fuels. Nowadays, a plan for the management of spent fuel is being sought focusing on a non-aqueous process in Korea. Named as pyroprocessing, it includes an electrolytic reduction process using molten salt at high temperature for the spent fuels, which provides metallic product for a following electro-refining process. The electrolytic reduction process utilizes electrochemical reaction producing Li to convert oxides into metals in high temperature LiCl medium. Various kinds of elements in the spent fuels would be distributed in the system according to their respective reactivity with the reductant, Li, and the medium, LiCl. This study elucidates the reactions of the elements to understand the behavior of composite elements on the spent fuels by thermodynamic calculations. Uranium and transuranic are reduced into their metallic forms while rare-earth oxides, except for Eu, are stable against the reaction at a process temperature. This study also covers the tendency of reactions with respect to the temperature and, finally, estimates radioactivity and heat load on the distributed phases based on the reference spent fuel characteristics.

• Journal of Institute of Convergence Technology
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• v.4 no.2
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• pp.47-50
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• 2014

Voloxidation is a process for converting $UO_2$ into $U_3O_8$ while removing some volatile products in spent fuels (SF). Various oxidative gas conditions including air and mixture of Ar and $O_2$ could be adopted for the process. The gas flows into a reactor under high temperature ($>500^{\circ}C$) and components of SF are reacted with the gas. SF is composed of various components such as actinides, lanthanides, and alkali metals. Therefore, it is of significance to understand their behavior during the reactions for process development. However, due to the limit of available experiments, phase diagram analysis should be preceded. TPP diagram is constructed with respect to temperature-pressure-pressure. It shows a stable phase depending on partial pressures of gas components as well as temperature. In this work, we investigated TPP diagrams for actinides, lanthanides and other oxides to determine stable oxide forms under different gas conditions. The results would be used to set up a material balance under a pyroprocessing scheme of SF and compare the gas conditions for the optimization of fission products removal.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2002.05a
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• pp.98-102
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• 2002

As the nuclear power plant has been operated continuously and increased gradually, transportation and storage of spent fuel are seriously considered nowadays. The transportation cask which contains radioactive material needs to be inspected about structural safety. About safety verification, description of IAEA Safety Standards states that cask must withstand hypothetical accident conditions. In this paper, 9m free drop impact analysis was performed for transportation cask and impact limiter by using the finite element methods. Furthermore, we obtained the dynamic behavior of wood to as compared with safety test results, and verified the safety of transportation cask.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.153-154
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• 2017

The thermal analysis of transportation and storage cask for SNF was conducted during short term loading operations for forced gas drying condition. The fuel cladding temperature in 6 regions of SNF in the cask during the short term loading operations for forced gas drying condition is shown in the Fig. 3. The thermal analysis results of calculated maximum cladding temperature in each process demonstrate that operating scenario of TFD in detailed design maintain well below the temperature limits of $400^{\circ}C$.