• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.223-232
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• 2018

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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.83-89
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• 2003

A temperature- and pressure-reducing process is utilized to handle the spent fuel assembly in the post-irradiation examination facility. This process includes three separated unit processes. First one is the decontamination process to clean the spent fuel assembly casks. The second process is the temperature-reducing process to reduce the temperature elevated by decay process in the spent fuel assembly. The third process is the filtration process to remove insoluble particles existed in the casks using filters. Up-to-date technologies as well as practical theories related to the temperature- and pressure-reducing process is reviewed in this report. The test-operation process for various tests and the test results of the temperature- and pressure-reducing process for J-44 and K-23 spent fuel assemblies are also described in detail. This report must be effectively used for the normal operation of the facility with the awareness of unprecedented problems which could occur by continuing operation of the PIE facility.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.627-631
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• 2003

In a various regulations and standards related to the spent fuel storage, the storage casks should be designed to sustain the structural integrity under the accident conditions of predicted operation and design criteria. These conditions for the structural evaluation requires the drop, tip-over, wind like tornado and typhoon, flood and earthquake. This paper describes the load cases and conceptual evaluation method for the structural evaluation. Preliminary safety analysis of the concrete storage system were peformed.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.248-254
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• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.135-149
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• 2017

Demonstration programs for spent nuclear fuel dry storage have been carried out to produce important and confirmatory data to support safety of dry storage systems and integrity of spent nuclear fuel stored in dry condition. The US initiated the dry storage of spent nuclear fuel and has strict and explicit regulatory stipulations on the integrity of spent nuclear fuel in dry storage. The US has carried out several notable demonstration programs for the initiation and license extension of dry storage. At the very early stage of dry storage, the demonstration programs were focused on proof of the safety of dry storage systems and a demonstration project called the dry cask storage characterization project was performed for the license extension of low burn-up fuel dry storage. Currently, a demonstration program for the license extension of high burn-up fuel dry storage is under way and is expected to continue for at least 10 years. Korea has not yet begun the dry storage of PWR fuel and the US programs can be a good reference and can provide lessons to safely begin and operate dry storage in Korea. In this paper, past and current demonstration programs of the US are analyzed and several recommendations are provided for demonstration programs for the dry storage of spent nuclear fuel in Korea.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.582-589
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• 2010

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.349-350
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• 2006

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.351-352
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• 2006