• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.117-121
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• 2016

Background: Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Materials and Methods: Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. Results and Discussion: The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. Conclusion: The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.301-313
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• 2015

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.359-372
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• 2005

The strong protection method of radiation emergency preparedness is the evacuation when a great deal of radionuclide material is released to environment. Required factors for evacuation time estimate of Ulchin nuclear power plant site were investigated. The traffic capacity and the traffic volume by season were investigated for the traffic analysis and simulation within EPZ of Ulchin site. As a result, the background traffic volume by season were established. The NETSIM code was applied to simulate for 12 scenarios in the event of normal traffic/summer peak traffic/winter peak traffic, daytime/night, and normal weather/adverse weather conditions. The results showed that the evacuation time required for total vehicles to move out from EPZ took generally $210\~315$ minutes. The evacuation time took longer about 45 minutes at night than in the daytime, and 45 minutes in adverse weather than normal condition.

• Mass Spectrometry Letters
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• v.2 no.1
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• pp.1-7
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• 2011

China's rapid economic growth has resulted in significant environmental side effects. Therefore, China has been interested in reducing her dependence on foreign oil and gas by developing technologies needed for hydrogen, in addition to her increasing energy mix of nuclear and renewable energy form, such as solar and wind power. There are three isotopes of hydrogen, i.e. protium (P or H), deuterium (D), and tritium (T). Both deuterium and tritium are important materials in nuclear fuel cycle industry. Tritium is one of the critical radioactive nuclides. Planning for and implementing contamination control as a part of normal operation and maintenance activities is an important function in any hydrogen facility, especially tritium facility. The development of hydrogen isotopes analysis is the key issues in this area. Mass spectrometry (MS) with medium (about 600) and high resolution (> 1,400) is commercially available; however, the routine analysis of hydrogen isotopes is done with low-resolution MS (< 200) in China. This paper summarizes the progress of MS measurement technology for hydrogen isotope abundance in China, focusing on our lab's research program and technical status. An analyzing method has been introduced for accurate measurement of tritium abundance in the H.D.T system by low resolution MAT-253 MS. The quotient of compression ratio coefficient is determined by building up equipment for laboratory-scale preparation of secondary standard gases and by considering the difference in sensitivity between hydrogen isotopes. The results show that the measured value is reproducible within the relative error range of 0.8% for gas samples of different tritium abundance.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.31-40
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• 2005

High temperature - high pressure apparatus was developed to simulate nickel fewite corrosion products which were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 $\^{circ}C$ in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needlelike or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.45-63
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• 2017

The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.

• 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.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.21-26
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• 2012

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at $1170^{\circ}C$ after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2~3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.