In order to use nuclear energy stably, high level radioactive waste including spent nuclear fuel that is inevitably discharged from nuclear power plants after electricity generation must be managed safely and isolated from the human living area for a long period of time. In consideration of the accumulated amount of spent nuclear fuel anticipated according to the national policy for HLW management, the area required for the deep geological repository facility is expected to be very large. Therefore, it is essential to conduct various studies to optimize the area required for the disposal of spent nuclear fuel in cases where the nationally available land is extremely limited, such as in Korea. In this study, as part of such research, the strategies and the requirements for the preliminary design of a high efficiency repository concept of spent nuclear fuel were established. For PWR spent nuclear fuel, seven assemblies of spent nuclear fuel can be accommodated in a disposal canister, and high burnup of spent nuclear fuel was taken into consideration, and the source terms such as the amount and time of discharge and disposal were based on the 2nd national basic plan. By evaluating the characteristics, the amount of decay heat that can be accommodated in the disposal canister was optimized through the combination of seven assemblies of spent nuclear fuel. The cooling period of the radiation source for the safety assessment of the repository system was set at 55 years, and the operation of the repository would start from 2070 and then the disposal schedule would be conducted according to the disposal scenario based on the national basic plan. With these disposal strategies described above, the main requirements for setting up the conceptual design of the high efficiency repository system to be carried out in this study were described below. • A combination of seven spent nuclear fuels with high heat and spent nuclear fuels with low heat was loaded into a disposal canister, and the thermal limit per disposal canister was 1,600 W. • In order to maintain the long-term performance of the repository, the maximum temperature design limit in the buffer material was set to 130°C. • In the deep disposal environment, the safety factor [yield strength/maximum stress] required to maintain the structural stability of the disposal canister should be maintained at 2.0 or higher so that integrity of the canister can be maintained even under long-term hydrostatic pressure and buffer swelling pressure in the deep disposal environment. • The repository should have a maximum exposure dose of 10 mSv/yr or less, which is the legal limit in case of a single event such as an earthquake, and the risk level considering natural phenomena and human intrusion, which is less than the legal limit of 10-6/yr. These strategies and requirements can be used to develop the high-efficiency geological disposal concept for spent nuclear fuels as an alternative disposal concept.

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

The research for the safe management of high-level waste in Korea has been conducted by the Korea Atomic Energy Research Institute since 1997, and the results have formed the basis of the national basic plan for the high-level waste management and the revised national basic plan. In the future, it is evolving and developing R&D focusing on securing technologies for demonstration of the disposal technologies and R&D to develop disposal concepts that increase safety and improve efficiency. Efficient management of heat generated from high-level radioactive waste, including spent nuclear fuel, is an important factor in establishing the disposal concepts because it must be in harmony with key factors such as repository layout, waste disposal container specifications, and design and operation for the barriers of the disposal system. For safe and complete isolation of highlevel radioactive waste in the deep geology, the disposal systems that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. These disposal systems were based on low burn-up spent nuclear fuel characteristics generated in the early stages of nuclear power generation, and next, based on the high-level wastes from recycling process of the high burn-up spent nuclear fuels, and were the direct disposal systems for the high burn-up spent nuclear fuels. So, it is necessary to track and analyze the change process in the decay heat characteristics of the high-level waste to be disposed of in order to improve the disposal concept, which enhances the safety of disposal and the utilization of the national land. Therefore, in this paper, the process of change in decay heat of reference spent nuclear fuels for disposal applied to the disposal concepts from the initial stage of development of high-level waste disposal technology to the present in Korea is analyzed.

In this paper, an approach developed by the Finnish nuclear waste management organization, Posiva, for the construction license of a geological repository was reviewed. Furthermore, a computer program based on the approach was developed. By using the computer program, the lifetime of a copper disposal canister, which was a key engineered barrier of the geological repository, was predicted under the KAERI Underground Research Tunnel (KURT) geologic conditions. The computer program was developed considering the mass transport of corroding agents, such as oxygen and sulfide, through the buffer and backfill. Shortly after the closure of the repository, the corrosion depths of a copper canister due to oxygen in the pores of the buffer and backfill were calculated. Additionally, the long-term corrosion of a copper canister due to sulfide was analyzed in two cases: intact buffer and eroded buffer. Under various conditions of the engineered barrier, the corrosion lifetimes of the copper canister due to sulfide significantly exceeded one million years. Finally, this study shows that it is necessary to carefully characterize the transmissivity of rock and sulfide concentration during site characterization to accurately predict the canister lifetime.

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS+). The results confirmed that the calculated effective multiplication factors (keff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy‧hr-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS+ system without any shielding issues. The container thickness applied to the current KRS+ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS+ design is suitable in terms of nuclear criticality and radiation shielding.

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

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%.

현재 기준개념으로 개발하여 상용화 단계에 있는 심층 동굴 처분기술에 대한 대안으로서 지질학적 조건이 더 안정적인 지하 3~5 km의 심도에 사용후핵연료를 포함한 고준위폐기물을 처분하는 심부시추공 처분기술의 국내 적용 가능성을 예비 평가 하였다. 이를 위하여 심부시추공 처분개념의 기술적 적용성 분석에 필요한 국내 기반암 분포특성 및 심부시추공 처분부 지적합성 평가 기술 분석과 대구경 심부시추기술을 평가하였다. 이들 분석결과를 바탕으로 심부시추공 처분시스템 설계 기준 및 요건에 적합한 심부시추공 처분용기 및 밀봉시스템 개념을 설정하여 예비 기준 심부시추공 처분 개념을 도출하였다. 그리고 도출된 예비 기준 처분시스템에 대하여 열적 안정성 및 그래픽 처분환경에서의 처분공정 모사 등 다양한 성능평가를 수행하고 이들을 종합하여 심부시추공 처분시스템의 국내 적용성에 대하여 다양한 관점에서의 예비평가를 수행하였다. 결론적으로, 심부시추공 처분시스템은 처분심도와 단순한 방법으로 인하여 안전성 및 경제적 타당성 측면에서 많은 장점이 있지만, 불확실성을 줄이고 인허가를 획득하기 위해서는 이 기술에 대한 현장실증이 필수적이다. 본 연구결과는 사용후핵연료 관리 국가정책 수립을 위한 공학적 근거자료로 활용이 가능하며, 심부시추공 처분기술에 관심을 갖는 방사성폐기물 관리 이해당사자들에게 필요한 정보자료로 제공될 수 있다.

본 연구에서는 심부시추공 처분을 위한 밀봉시스템으로서 Gibb’s Group에 의해 제안된 화강암 용융 및 재결정화에 의한 시 추공 밀봉 방안에 대해 KURT 화강암을 대상으로 실현 가능성을 확인하였다. 화강암 용융 실험은 첨가제를 이용한 상압용 융시험과 물의 기화에 의한 수증기 고압용융시험 2가지로 수행되었다. 상압 용융시험 결과, KURT 화강암 분말에 NaOH를 첨가하여도 기본 융점보다 낮은 1,000℃에서 부분용융이 시작되었으며, 냉각된 용융물에서 침상결정의 형성을 확인하였다. 수증기 고압시험은 물의 첨가량에 따라 수증기압을 달리하며 최대 400 bar의 수증기압까지 용융 시험이 진행되었다. KURT 화강암은 낮은 수증기압에도 1,000℃에서 부분 용융이 시작되었으나, 물이 많이 첨가된 높은 수증기압에서 화강암의 부분 용융은 보이지 않았다. 따라서 소량의 수증기가 있는 고압상태가 화강암의 용융에 적합한 것으로 판단되었다. 한편, 고온고압의 수증기는 내부식성의 반응기 벽을 부식시켜, 고온의 수증기에 의한 처분용기의 부식 문제가 발생되었다.

고준위 방사성폐기물로 분류되는 사용후핵연료를 현재 기술로 가장 안전한 격리 방법으로는 500 m 심도의 안정한 암반에 심지층 처분하는 방법으로, 가장 중요한 요건은 공학적방벽인 완충재의 온도가 100℃를 초과하지 않도록 시스템을 설계하 는 것이다. 국내의 경우 전체 전력 소요량의 약 30% 정도를 차지하고 있는 원자력발전으로 발생되는 사용후핵연료의 양은 지속적으로 증가하여 누적되고 있어, 이들을 처분하기 위한 소요면적도 증가하고 있다. 따라서, 본 연구에서는 처분면적을 감소시킴으로써 처분효율을 향상시키기 위한 목적으로 다양한 복층처분 개념을 도출하였다. 이를 바탕으로 중요한 처분시 스템 요건 만족여부를 확인하기 위하여 열해석을 수행하고 그 결과를 분석하여 처분시스템 열적 안정성을 평가하였다. 평 가결과, 기준시스템 위치인 500 m 심도로부터 상부 또는 하부로 75 m를 이격한 심도에 복층으로 처분시스템 구축이 가능 하였으며, 실제 부지특성자료에 따른 상세 분석이 요구된다. 본 연구결과는 사용후핵연료 관리정책 수립 및 실제 처분시스 템 설계에 활용될 것으로 사료된다.

사용후핵연료 또는 고준위폐기물의 안전한 처분을 위하여 지난 수십 년 동안 많은 나라들이 다양한 처분대안을 연구하여 왔다. 본 논문에서는 심지층처분기술에 있어서 사용후핵연료를 직접 처분하는 방안으로서 처분효율 향상을 위한 다양한 방 안 중의 하나로 고려할 수 있는 PWR 사용후핵연료 집합체를 해체하여 연료봉을 밀집한 경우에 대한 처분 효율을 분석하였 다. 이를 위하여, 우선 사용후핵연료 연료봉 밀집개념과 관련 처분용기 및 심지층처분 개념을 설정하였다. 이 개념에 근거하 여 심지층 처분시스템의 공학적방벽 설계에 있어서 가장 중요한 요건인 완충재의 온도 제한요건을 만족시키는지 여부를 확 인하기 위하여 각 처분개념 별로 열해석을 수행하였다. 그리고, 처분공 간격, 처분터널 간격 및 처분용기 열발산 면적에 따 른 열해석 결과를 바탕으로, 단위처분면적 관점에서의 처분효율을 비교/분석하고 평가하였다. 또한, 사용후핵연료봉을 밀 집시킨 경우에 있어서 냉각기간에 따른 처분개념을 분석하였다. 분석결과에 따르면 사용후핵연료봉을 밀집하여 심지층처 분하는 경우 처분효율 측면에서 불리한 것으로 판단되었다. 다만, 사용후핵연료의 냉각기간을 70년 이상으로 장기화 할 경 우 처분효율은 향상될 것으로 예상되지만, 사용후핵연료의 내구성 및 장기저장에 따른 조건 등 추가적인 분석이 필요하다.

사용후핵연료를 포함하는 고준위 방사성폐기물을 지질학적 조건이 안정적인 지하 3~5 km의 심도에 처분할 수 있다면 다음 과 같은 많은 장점이 있는 것으로 평가되고 있다. 즉, (1)암반 수리전도도가 매우 낮아 지하수가 생태계까지 도달하는데 속 도가 현저히 감소되며, (2)상부층 두께로 인하여 생태계와의 이격거리 확보에 유리하고, (3)지하수가 환원상태이므로 핵종 의 용해도가 매우 낮을 뿐만 아니라 (4)오랜 연령의 지하수에서는 핵종이 흡착된 콜로이드 생성과 이동이 극히 제한된다는 점이다. 이와 관련하여 심부시추공 처분(Deep Borehole Disposal) 연구는 심층 처분(Deep Geological Disposal) 시스템에 대한 이상적인 처분 대안기술로서 꾸준하게 진행되어 왔다. 본 논문에서는 최근 심부 시추기술이 비약적으로 발전됨에 따 라 의미있게 연구가 진행되고 있는 심부시추공 처분시스템을 국내 적용하기 위한 초기 단계로서 해외의 심부시추공 처분시 스템 기술개발 사례를 분석하였다. 이를 통하여 심부시추공 처분에 대한 일반적인 개념과 심부시추공 처분시스템 개념을 도 출한 연구사례를 국가별로 정리하였다. 이들 분석결과는 향후 심부시추공 처분기술의 국내 적용을 위한 입력자료로서 유용 하게 활용될 수 있을 것이다.