• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.873-880
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• 2017

To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2-23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to $100^{\circ}C$ and desorbs various organic and inorganic carbon compounds reaching to $200^{\circ}C$. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over $850^{\circ}C$, in order to reduce the C-14 concentration below the clearance level.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.279-298
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• 2019

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.229-236
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• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.199-212
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• 2007

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or dispersal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions. For the tests we examined the failure of bolts and the deformation of flange to evaluate a loss or dispersal of radioactive material and measured the shielding thickness using a ultrasonic thickness gauge to assess a loss of shielding integrity. The strains and accelerations acquired from tests were compared with those by analyses to verify the impact dynamic analysis. The analytic results were larger than the those of test so that the analysis showed the conservative results. Finally, we evaluated the safety of the type IP-2 transport package under the stacking test condition using a finite element analysis. Under the stacking test condition, the maximum Tresca stress of the shielding material was 1/3 of the yielding stress. Two kinds of a type IP-2 transport package were safe for the free drop test condition and the stacking test condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.285-299
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• 2006

At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.245-261
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• 2019

Radioactive contaminant from a nuclear facility moves to the ecosystem by run-off or groundwater flow. Among the two mechanisms, contaminant plume through a river can be easily detected through a surface water monitoring system, but radioactive contaminant transport in groundwater is difficult to monitor because of lack of information on flow path. To understand the contaminant flow in groundwater, understanding of the geo-environment is needed. We suggest a method to decide on monitoring location and points around an imaginary nuclear facility by using the results of site characterization in the study area. To decide the location of a monitoring well, groundwater flow modeling around the study area was conducted. The results show that, taking account of groundwater flow direction, the monitoring well should be located at the downstream area. Also, monitoring sections in the monitoring well were selected, points at which groundwater moves fast through the flow path. The method suggested in the study will be widely used to detect potential groundwater contamination in the field of oil storage caverns, pollution by agricultural use, as well as nuclear use facilities including nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.257-265
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• 2007

Spent filter wastes of about 1,000 units (200 L) have been stored in the waste storage facility of the Korea Atomic Energy Research Institute since its operation. At the moment, to secure space in a waste storage facility as well as to efficiently manage spent filter wastes, it is necessary to conduct a compaction treatment of these spent filters, and finally, to repack the compacted spent filters into a 200 liter drum. To do that, the spent filter wastes were first classified according to their generation facilities, their generation date and their surface dose rate by investigating the inventory of the spent filters. In order to repack a compacted spent filter in a 200 liter drum, it is first necessary to conduct a radionuclide assessment of a spent filter before compacting it. Therefore, after taking a representative sample from a spent filter without a dismantlement, the nuclide analysis for it will be conducted. And then, after putting a spent filter into a regular drum by conducting the columnar shaping of the hexahedral form of a spent filter, the compaction treatment of the shaped spent filter will be conducted by vertically compacting it.

• Computational Structural Engineering
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• v.6 no.4
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• pp.89-98
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• 1993

As the construction of nuclear power plants are increased, nuclear wastes disposal has been faced as a serious problem. If nuclear wastes are to be buried in the underground stratum, thermo-mechanical behavior of stratum must be analyzed, because high temperature distribution has a significant effect on tunnel and surrounding stratum. In this study, in order to analyze the structural behavior of the underground which is subject to concentrated heat sources, a coupling method of nonlinear finite elements and linear boundary elements is proposed. The nonlinear finite elements (NFE) are applied in the vicinity of nuclear depository where thermo-mechanical stress is concentrated. The boundary elements are also used in infinite domain where linear behavior is expected. Using the similar method as for the problem in mechanical field, the coupled nonlinear finite element-boundary element (NFEBE) is developed. It is found that NFEBE method is more efficient than NFE which considers nonlinearity in the whole domain for the nuclear wastes depository that is expected to exhibit local nonlinearity behavior. The effect of coefficients of the rock mass such as Poisson's ratio, elastic modulus, thermal diffusivity and thermal expansion coefficient is investigated through the developed method. As a result, it is revealed that the displacements around tunnel are largely dependent on the thermal expansion coefficients.

• Journal of the Korean institute of surface engineering
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• v.34 no.1
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• pp.62-67
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• 2001

Green technology is being developed up to a point that is feasible not only in an environmental sense, but also in an economical viewpoint. This paper introduces two case studies that applied $CO_2$ technology into nuclear industry. 1) Nuclear laundry : A laundry machine that uses liquid and supercritical $CO_2$ as a solvent for decontamination of contaminated working dresses in nuclear power plants was developed. The machine consists of a 16 liter reactor, a recovery system with compressors, and storage tanks. All $CO_2$ used in cleaning is fully recovered and reused in next cleaning, resulting in no production of secondary nuclear waste. Decontamination factor is still lower than that in the methods currently used in the plant. Nuclear laundry using $CO_2$ looks promising with technical improvements-surfactants and mechanical agitation. 2) $CO_2$ nozzle decontamination : An adjustable nozzle for controlling the size of dry ice snow was developed. Using the developed nozzle, a surface decontamination device was made. Human oils like fingerprints on glass were easy to remove. Decontamination ability was tested using a contaminated pump-housing surface. About 40 to 80% of radioactivity was removed. This device is effective in surface-decontamination of any electrical devices like detector, controllers which cannot be cleaned in aqueous solution.