• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.333-349
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• 2013

As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safety-criteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.243-251
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• 2015

Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.175-182
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• 2009

The spent cationic exchange resins and anionic exchange resins were separated from mixed spent exchange resins by a fluidized bed gravimetric separator. The separated resins were identified by an elemental analysis and thermogravimetric analysis. The each test sample was prepared by diluting the slurry made by wet ball milling the cationic exchange resins and the anionic exchange resins separated as a spherical granular form for 24 hours. The resulting test samples showed a slurry form of less than $75{\mu}m$ of particle size and 25,000ppm of $COD_{cr}$. The decomposition conditions of each test samples from a thermal power plant were obtained with a lab-scale(reactor volume : 220mL) supercritical water oxidation(SCWO) facility. Then pilot plant(reactor volume : 24 L) tests were performed with the test samples from a thermal power plant and a nuclear power plant successively. Based on the optimal decomposition conditions and the operation experiences by lab-scale facility and the pilot plant, a commercial plant(capacity : 150kg/h) can be installed in a nuclear power plant was designed.

• Journal of the Mineralogical Society of Korea
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• v.26 no.2
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• pp.119-127
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• 2013

Borosilicate glass dissolution is an important chemical process that impacts the glass durability as nuclear waste form that may be used for high-level radioactive waste disposal. Experiments reported that the glass dissolution rates are strongly dependent on the bulk composition. Because some relationship exists between glass composition and molecular-structure distribution (e.g., non-bridging oxygen content of $SiO_4$ unit and averaged coordination number of B), the composition-dependent dissolution rates are attributed to the bulk structural changes corresponding to the compositional variation. We examined Na borosilicate glass structures by performing classical molecular dynamics (MD) simulations for four different chemical compositions ($xNa_2O{\cdot}B_2O_3{\cdot}ySiO_2$). Our MD simulations demonstrate that glass surfaces have significantly different chemical compositions and structures from the bulk glasses. Because glass surfaces forming an interface with solution are most likely the first dissolution-reaction occurring areas, the current MD result simply that composition-dependent glass dissolution behaviors should be understood by surface structural change upon the chemical composition change.