KORAD (Korea Radioactive Waste Agency, http://www.korad.or.kr) has stored slightly contaminated ascon (asphalt coated concrete mixture) that was introduced to Gyeongju repository about a decade ago waiting for a final disposal. It is believed to be mainly contaminated by radioisotope 137Cs due to impurities introduced from the outside during the ascon manufacturing process. We studied characteristics of the radioactive waste to see whether this material would be proper enough to be disposed in Gyeongju LILW repository or be other ways to reduce the disposal volume including self-disposal before its final disposal otherwise. KORAD looked into the properness of characteristics of ascon in terms of WAC (Waste Acceptance Criteria) documented by KORAD that includes general chemical and physical properties of asphalt, density, size of grains, content of organic material and possibility of existence of chelate materials that qualitatively limited to be disposed by the criteria. And other associated characteristics such as gas generation and bio degradation were also investigated. Based on the data obtained from the study, we proposed various plausible solutions in associated with operational and disposal safety and economic view points. This study will be used for KORAD’s decision on how to control and safely dispose the spent ascon within a reasonable time period. And also those experiences may be applied for other LILW issues that require treatment or conditioning of radioactive wastes in the future.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

Domestic NPPs had produced the paraffin-solidifying concentrate waste (PSCW) for nearly 20 years. At that time radioactive waste management policy of KHNP was to reduce the volume and to store safely in site. The PSCW has been identified not to meet the leaching index after introducing the treatment system. PSCW has to be treated to meet current waste acceptance criteria (WAC) for permanent disposal. PSCW consists of dried concentrate 75% and paraffin 25% of volume. When PSCW is separated into a dried concentrate and a paraffin by solubility, total volume separated is increased twice. Final disposal volume of dried concentrate can reach to several times when solidifying by cement even considering exemption. Application of polymer solidification technology is difficult because dried concentrate is hard to make form to pellet. When PSCW is packaged in High Integrity Container (HIC), volume of PSCW is equal to the volume before package. The packaging process of HIC is simple and is no necessary of large equipment. It is important to recognize that HIC was developed to replace solidification of waste. HIC has as design goal a minimum lifetime of 300 years under disposal environment. The HIC is designed to maintain its structural integrity over this period, to consider the corrosive and chemical effects of both the waste contents and the disposal environment, to have sufficient mechanical strength to withstand loads on the container and to be capable of meeting the requirements for a Type A transport Package. The Final waste form is required for facilitating handling and providing protection of personnel in relation to solidification, explosive decomposition, toxic gases, hazardous material, etc. Structural stability of final waste form is required also. Structural stability of the waste can be provided by the waste itself, solidifying or placing in HIC. Final waste form ensure that the waste does not structurally degrade and affect overall stability of the disposal site. The HIC package contained PSCW was reviewed from several points of view such as physicochemical, radiological and structural safety according to domestic WAC. The result of reviewing shows that it has not found any violation of WCP established for silo type disposal facility in Gyeongju city.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.

국내 3단계 매립형 처분시설은 2018년도 한국원자력환경공단의 중^저준위 방폐물관리시행계획에 의하면 주로 원전 해체 현장에서 발생하는 극저준위방폐물을 수용하기 위해 2019년 4월부터 2026년 2월까지 총 104,000드럼(2개 트렌치)을 수용 하기 위해 건설이 계획 중이다(총 2,246억원 투입). 이후 총 5개 트렌치에 260,000드럼이 총 34,076 m2의 면적에 단계적으로 수용되며 따라서 현재 한국원자력환경공단은 관련 인수기준을 마련 중에 있다. 극저준위방폐물 처분시설 인수기준의 경우 프랑스, 스페인 등이 전용 처분시설을 운영하면서 자국의 인수기준을 합리적으로 잘 준용하고 있으나 본 논문에서는 해체방 폐물의 처분에 가장 경험이 많은 미국의 처분시설을 고려하여 국내 매립형 처분시설에 우선적으로 반영되어야 할 사항이 있는지 분석하였고 이를 통하여 경주내 3단계 매립형 처분시설의 인수기준 마련에 도움이 되고자 하였다.

원자력발전을 지속가능한 에너지원으로 활용하기 위해서는 원전 해체 및 운영 과정에서 발생하는 방사성폐기물의 안전하고 효율적인 처분이 매우 중요하다. 방사성폐기물 종류는 다양하지만 해체과정에서 가장 많이 발생할 것으로 예상되는 극저준 위방사성폐기물 인수기준수립은 원전해체전략수립에 큰 영향을 줄 것으로 보인다. 본 연구에서는 영국과 미국의 극저준위 방사성폐기물처분장 인수기준을 경주에 건설 중인 원자력환경센터의 인수기준과 비교분석을 통해 향후 우리나라 극저준위 방사성폐기물 처분을 위한 폐기물 인수기준을 분석하고자 한다.

가새형 소성 감쇠기는 에너지 소산 이력거동을 통해 강한 지진하중을 받는 구조물의 구조적 손상을 방지하거나 감소시킨다. 본 연구에서는 성능수준 만족을 위한 가새형 소성 금비기의 직접적인 설계 방법을 개발하였다. 많은 해석 시간이 요구되는 비선형 동적 시간이력해석 대신 비선형 정적해석법인 능력스펙트럼법을 이용하여 주어진 성능을 만족하기 위하여 필요한 유효 감쇠비를 구한 후 이를 이용하여 가새형 소성 감쇠기의 크기를 구하였다. 각 설계변수의 영향을 파악하기 위하여 단자유도계에서 구조물의 주기, 요구되는 탄성강도에 대한 항복강도의 비, 항목 후 강성비, 가새형 소성 감쇠기의 항복응력 등을 변수로 하여 해석을 수행하였다. 본 연구를 통해 제안된 방법을 5층과 10층 건물에 적용하여 검증하였다. 시간이력해석 결과, 제안된 방법에 따라 설계된 가새형 소성 감쇠기를 설치한 예제 구조물의 최대응답은 주어진 목표변위와 잘 일치하였다.