Even though a huge amount of spent nuclear fuels are accumulated at each nuclear power plant site in Korea, our government has not yet started to select a final disposal site, which might require more than several km2 surface area. According to the second national plan for the management of high-level radioactive waste, the reference geological disposal concept followed the Finnish concept based on KBS-3 type. However, the second national plan also mentioned that it was necessary to develop the technical alternatives. Considering the limited area of the Korean peninsula, the authors had developed an alternative disposal concepts for spent nuclear fuels in order to enhance the disposal density since 2021. Among ten disposal concepts shown in the literature published in 2000’s, we narrowed them to four concepts by international experiences and expert judgements. Assuming 10,000 t of CANDU spent nuclear fuels (SNF), we designed the engineered barriers for each alternative disposal concept. That is, using a KURT geological conditions, the engineered barrier systems (EBS) for the following four alternative concepts were proposed: ① mined deep borehole matrix, ② sub-seabed disposal, ③ deep borehole disposal, and ④ multi-level dispoal. The quantitative data of each design such as foot prints, safety factors, economical factors are produced from the conceptual designs of the engineered barriers. Five evaluation criteria (public acceptance, safety, cost, technology readiness level, environmental friendliness) were chosen for the comparison of alternatives, and supporting indicators that can be evaluated quantitatively were derived. The AHP with domestic experts was applied to the comparison of alternatives. The twolevel disposal was proposed as the most appropriate alternative for the enhancement of disposal efficiency by the experts. If perspectives changes, the other alternatives would be preferred. Three kinds of the two-level disposal of CANDU SNF were compared. It was decided to dispose of all the CANDU spent nuclear fuels into the disposal holes in the lower-level disposal tunnels because total footprint of the disposal system for CANDU SNF was much smaller than that for PWR SNF. Currently, we reviewed the performance criteria related to the disposal canister and the buffer and designed the EBS for CANDU SNF. With the design, safety assessment and cost estimates for the alternative disposal system will be carried out next year.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

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%.

후행 핵연료주기 경제성 평가는 추정 비용의 불확실성, 평가 대상기간의 장기성, 적용 할인율에 따른 계산결과의 변동성 등 많은 불확실성을 내포하고 있기 때문에 평가기관 또는 평가자에 따라 그 결과가 서로 상이하다. 본고에서는 지금까지 수행 된 주요 경제성 평가 연구들을 조사/분석하여 그 특징과 한계를 알아봄으로써 현재 국내에서 추진되고 있는 사용후핵연료 공론화 및 후행 핵연료주기 정책 연구 추진에 기초자료로 활용될 수 있도록 하고자 하였다. 분석 결과 사용후핵연료 재활용 옵션에 비해 직접처분 옵션이 유리하나, 입력 자료로 사용된 파라미터 값에 따라 결과의 불확실성이 많이 나타나 이 부분에 대한 추가적인 연구가 필요하다는 사실을 알 수 있었다.

제2차 전력수급기본계획에 의거 2017년까지 계획된 원자로만을 대상으로 심지층 처분시스템 설계 시 필요한 국내 사용후핵연료의 발생량, 제원적 특징, 초기농축도 및 방출연소도 등에 대하여 현재 및 미래 현황을 파악하고 예측하였다. 2057년까지 PWR 및 CANDU 사용후핵연료 발생량은 각각 20,500 및 14,800 MTU로 나타났다. 초기 농축도에 대해서는 4.5 wt.% 이하를 갖는 사용후핵연료가 96.5%를 차지하는 것으로 나타났다. 사용후핵연료의 평균 방출연소도는 90년대 후반에는 36 GWD/MUT 전도, 2000년대 초반에는 40 GWD/MTU를 나타냈으며, 2000년대 중 후반부터는 45 GWD/MTU가 될 것으로 나타났다. 광범위한 분석 및 예측 결과, 총 처분물량을 대표할 수 있는 가상적인 기준 사용후핵 연료는 16 6 한국표준형연료, 단면적 , 길이 453cm, 무게 672 kg, 초기 농축도 4.5 wt.%, 방출연소도 55 GWD/MTU로 나타났다.