• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.115-124
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• 2013

Among the several options to solve PWR spent fuel accumulation problem in Korea, the dry storage method could be the most realistic and applicable solution in the near future. As the basic objectives of dry storage are to prevent a gross rupture of spent fuel during operation and to keep its retrievability until transportation, at the same time the importance of a spent fuel integrity evaluation that can estimate its condition at the final stage of dry storage is very high. According to the national need and technology progress, two representative nations of spent fuel dry storage, the USA and Japan, have established different system temperature criteria, which is the only controllable factor in a dry storage system. However, there are no technical criteria for this evaluation in Korea yet, it is necessary to review the previously well-organized methodologies of advanced countries and to set up our own domestic evaluation direction due to the nation's need for dry storage. To satisfy this necessity, building a domestic spent fuel test database should be the first step. Based on those data, it is highly recommended to compare domestic data range with foreign results, to build our own criteria, and to expand on evaluation work into recently issued integrity problems by using a comprehensive integrity evaluation code.

• Analytical Science and Technology
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• v.18 no.4
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• pp.298-308
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• 2005

The methods for determining C-14 and tritium contents in the spent nuclear fuel sample were developed. The carbon-14($^{14}CO_2$) released during the dissolution of the spent fuel sample and $CaCO_3$ ($CO_2$ carrier) with 8 M $HNO_3$ at $90^{\circ}C$ was collected in trap containing 1.5 M NaOH. The volatile radioactive iodine evolved when the spent fuel was dissolved, was trapped on to Ag-silicagel (Ag-impregnated silicagel) adsorbent in column which is connected to two NaOH traps. The solutions which contain tritium as HTO after fuel dissolution were decontaminated by deionization with a mixture of cation and anion exchange resins and inorganic ionexchangers. The amount of C-14 in the trap solutions and the HTO concentration in the resulting deionization water were then determined by liquid scintillation counting.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1997.10a
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• pp.617-620
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• 1997

In this paper, the thermal stress analysis of spent nuclear fuel disposal canister in a deep repository at 500m underground is done for the underground pressure variation. Since the nuclear fuel disposal usually emits much heat and radiation, its careful treatment is required. And so a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences some mechanical external loads such as hydrostatic pressure of underground water, swelling pressure of bentonite buffer, and the thermal load due to the heat generation of spent nuclear fuel in the basket etc.. Hence, the canister should be designed to designed to withstand these loads. In this paper, the thermal stress analysis is done using the finite element analysis code, NISA.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.11a
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• pp.338-344
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• 2003

The seismic analysis method of spent fuel storage structures for PHWR plant is introduced in comparison with the method for PWR plant. Investigating the structural characteristics of the storage structures, the former is vertically stacked fuel storage trays, while the latter is welded honeycomb type structure. However, as both structures are submerged and free standing, the analysis methods to anticipate the seismic response of both structures are complicated. For the better estimation of actual seismic response, how to model the dynamic properties and the structural behaviour is the key issue. In this paper, the overall procedures of the seismic modelling and stability check for seismic sliding and overturning of the two different storage structures are discussed in the viewpoint of analysis reliability

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.1-7
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• 2009

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.221-222
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.362-363
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• 2017

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.383.1-383
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.312-312
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• 1998

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.