• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.427-438
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• 2013

In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, will not suffice to handle all the spent fuel discharged annually. Hence centralized dry cask storage will be required even if the pyroprocessing option is successfully developed by 2030. Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR). We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-fromreactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.383-389
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• 2005

This research is to develop a seismic response analysis method for a spent fuel storage cask. FEM model is built for the test model of 1/8 scale spent fuel dry storage cask using available 3D contact conditions in ABAQUS/Explicit. Input load for this analysis os a seismic wave of El-centro earthquake, and the friction and damping coefficients in the analysis condition we obtained from the test result. Penalty and kinematic contact methods of ABAQUS are used for mechanical contact formulation. The analysis method was verified for rocking angle obtained by seismic response tests. The kinematic contact method with an adequate normal contact stiffness showed a good agreement with tests.

• Structural Engineering and Mechanics
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• v.75 no.2
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• pp.239-246
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• 2020

In this paper, a novel analytical method, Max-Min Approach (MMA), has been presented and applied to consider the nonlinear vibration of dry cask storage systems. The nonlinear governing equation of the structure has been developed using the shell theory. The MMA results are compared with numerical solutions derived by Runge-Kutta's Method (RKM). The results indicate a satisfying agreement between MMA and numerical solutions. Parametric studies have been conducted on the nonlinear frequency of dry casks. The phase-plan of the problem is also presented and discussed. The proposed approach can potentially ca be extended to highly nonlinear problems.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.91-99
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• 2017

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radio-active materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B&PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

• 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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.505-517
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• 2016

The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD) was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is freestanding on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact loade-time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

• 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 $400^{\circ}C$.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.345-345
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• 2004

The spent fuel transportation casks have used as one of the most essential component in the nuclear industry. And, the number of the cask has been significantly increased in recent years. While the bulk amount of spent fuel in the world is still kept in the storage pool, the number of countries which have chosen the advantages of dual purpose cask for transportation and storage is rapidly increasing. The technical experience in the area of spent fuel transportation cask operation and maintenance for long period is also available and will be well utilized also in storage casks. The increasing use of casks for dual and multiple purposes raises an issue of long term consideration by international standardization. Accordingly IAEA is providing a regulatory requirements and guidelines as an effort for this standardization.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.147-154
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• 2008

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.