• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.119-131
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• 2020

A numerical model was developed using COMSOL Multiphysics to evaluate groundwater flow that causes radionuclide migration in the unsaturated zone of a near-surface disposal facility, which is considered as a domestic low and an intermediate-level radioactive waste disposal facility. Each scenario was modeled by constructing a two-dimensional domain that included the disposal vault, backfill, disposal cover, and unsaturated aquifer. A comparison of the continuous and intermittent rainfall conditions exhibited no significant difference in any of the factors considered except the wave pattern of water saturation. The input data, such as porosity and residual water content of the unsaturated aquifer, were observed to not have a significant effect on the groundwater flow. However, the hydraulic conductivity of the unsaturated aquifer was found to have a significant effect on the groundwater flow. Therefore, it is necessary to assess the hydraulic conductivity of an unsaturated aquifer to determine the extent of groundwater infiltration into the disposal vault.

• Explosives and Blasting
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• v.35 no.3
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• pp.31-54
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• 2017

In the countries operating nuclear reactors, the development of high-level radioactive waste(HLW) disposal technique is considered as an urgent and important issue for sustainable utilization of nuclear energy. In Korea, in which a low and intermediate radioactive waste repository is already operating, the construction of an underground research laboratory for in situ validation studies became a matter of interest with increasing concerns on the management of HLW. In order to construct and to operate an underground HLW repository safely in deep underground, the stability of rock mass should be guaranteed. As an important factor on rock stability, excavation damaged zone (EDZ) has been studied in many underground research laboratories in foreign countries. For accurate evaluation of the characteristics and effects of EDZ under disposal condition, it is required to use reliable investigation method based on the analysis of previous studies in similar conditions. In this study, status of foreign underground research laboratories in other countries, approaches for investigation the characteristics, size, and effect of EDZ, and major findings from the researches were surveyed and reported. This will help the accomplishment of domestic researches for developing HLW management techniques in underground research laboratory.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.387-398
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• 2003

For reference human intrusion scenarios constructed in previous study, a probabilistic safety assessment to derive the radionuclide concentration limits for the low- and intermediate- level radioactive waste disposal facility is conducted. Statistical approach by the Latin Hypercube Sampling method is introduced and new assumptions about the disposal facility system are examined and discussed. In our previous study of deterministic approach, the post construction scenarios appeared as most limiting scenario to derive the radionuclide concentration limits. Whereas, in this statistical approach, the post drilling and the post construction scenarios are mutually competing for the scenario selection according to which radionuclides are more important in safety assessment context. Introduction of new assumption shows that the post drilling scenario can play an important role as the limiting scenario instead of the post-construction scenario. When we compare the concentration limits between the previous and this study, concentrations of radionuclides such as Nb-94, Cs-137 and alpha-emitting radionuclides show elevated values than the case of the previous study. Remaining radionuclides such as Sr-90, Tc-99 I-129, Ni-59 and Ni-63 show lower values than the case of the previous study.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.260-268
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• 1997

The effects of three mechanisms, calcium depletion, sulphate and carbonate penetration, on the concrete degradation have been studied. The shrinking core model (SCM) and the HYDROGEOC. HEM (HGC) model have been applied to evaluate how fast the mechanisms proceed. The SCM is an analytical approximation model and the HGC is a numerical mass transport model coupled with chemical reaction. The SCM leads to more conservative results than the HGC, and turns out to be very useful in the viewpoint of simplicity and conservatism. During 300 years, calcium has been depleted within 10 cm from the concrete outer surface, and sulphate has penetrated less than 13.5 cm into the concrete. Carbonate has not penetrated own 7 cm into the concrete in contact with the bentonite, and, furthermore, its penetration into the concrete with the groundwater is negligible. Conclusively, the concrete is expected to maintain its integrity for at least 300 years that are regarded as institutional control period of intermediate and low-level radioactive waste repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.117-123
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• 2012

In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.391-396
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• 1998

Modified product Consistency Test (M-PCT) has been developed as an alternative to other existing methods in determining the leachability of glass. M-PCT, the leaching method, is a hybrid of MCC-l and PCT, but can provide quicker sample preparation. Larger diameter glass sample (1.0-2.0 mm) than in the PCT method can be used so that the glass beads are more easily produced and cleaned. From the M-PCT, the total mass loss (ML) of glass, the normalized elemental release rate (NLi), pH value of leachate have been obtained. For some selected glasses in which leaching rates have been known, their chemical durablility have been tested using the M-PCT method. The results are compared to the literature data for the glasses. It is found that M-PCT method is reasonable and suitable in determining the leachability of Low and Intermediate level Radioactive Waste glass form, such as the pH, elemental loss and total mass loss.

• Tunnel and Underground Space
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• v.29 no.4
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• pp.215-229
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• 2019

This technical note describes the current status of Finnish radioactive waste disposal project which started to construct the repository for spent nuclear waste for the first time in the world. Finland started operating nuclear power plant in 1977 and is currently operating four nuclear power plants. After detailed site surveys started in 1993, Olkiluoto was finally selected by the parliament of Finland as the site for geological disposal in 2001 followed by a construction license in 2015. If the operating license is approved by the government in the 2020s, it would be the world's first case of geological disposal. In ONKALO, a site-specific underground research facility at the site of Olkiluoto, various studies were conducted to verify the safety of the repository. Finland uses the KBS-3 disposal concept, and Korea considers a similar disposal concept because of similar rock formations. The entire process in Finland including the operation status of intermediate and low-level waste disposal, site investigation and selection stages, and the latest rock mechanics and hydrogeological studies in ONKALO are presented. Suggestions for the radioactive waste disposal in Korea is given based on the Finnish case.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.357-368
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• 2008

On July 31, 2008, the Government issued the construction and operation permit for the first low and intermediate level radioactive waste disposal facility in the Republic of Korea. In this paper, the fundamental regulatory framework, regulatory requirements and technical standards of the disposal facility are introduced, and the phased review process adopted for evaluation of the safety of the facility is briefly described. The Atomic Energy Act sets forth a stepwise regulatory framework for the whole life-cycle of the disposal facility such as siting, design, construction, operation, closure and institutional control. More detailed regulatory requirements and technical standards are stipulated in the subsequent regulations of the Atomic Energy Act and a series of Notices issued by the Ministry of Eduction, Science and Technology. The Korea Institute of Nuclear Safety, as entrusted by the Ministry under the Atomic Energy Act, conducted safety review on the disposal facility, and evaluated the compliance with relevant criteria in all technical elements(i.e. siting and structural safety, radiological environmental impact, operational safety, systems and components, quality assurance, and total systematic performance assessment, etc.). The overall safety review process can be phased into inception phase, initial review phase, main review phase and completion phase. The review results were reported to and deliberated by the five Sub-committees of the Special Committee on Nuclear Safety, and then reported to the Ministry. The Ministry issued the construction and operation permit of the disposal facility through the deliberation of the review results by the Nuclear Safety Commission. Hereafter, the safety of the repository will be reassured by a series of subsequent regulatory inspections and reviews under the Atomic Energy Act. In addition, the licensee's continuous implementation of the "Safety Promotion Plan" may also enhance the long-term safety of the repository and contribute to build-up the confidence of the safety case.

• Journal of Radiation Protection and Research
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• v.45 no.4
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.437-445
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• 2019

After the permanent shutdown of K1 in 2017, decommissioning processes have attracted great attention. According to the current decommissioning roadmap, the dismantling of the activated components of K1 may start in 2026, following the removal of its spent fuel. Since the reactor vessel (RV) and reactor vessel internal (RVI) of K1 contain massive components and are relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. For achieving maximum efficiency of nuclear waste management processes for K1, we present activation analysis of the segmentation process and waste classification of the RV and RVI components of K1. For RVI, the active fuel regions and some parts of the upper and lower active regions are classified as intermediate-level waste (ILW), while other components are classified as low-level waste (LLW). Due to the RVI's complex structure and high activation, we suggest various underwater segmentation techniques which are expected to reduce radiation exposure and generate approximately nine ILW and nineteen very low level waste (VLLW)/LLW packages. For RV, the active fuel region and other components are classified as LLW, VLLW, and clearance waste (CW). In this case, we suggest in-situ remote segmentation in air, which is expected to generate approximately forty-two VLLW/LLW packages.