권호 표지

Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) (방사성폐기물학회지)

Korean Radioactive Waste Society

  • Quarterly
  • /
  • 1738-1894(pISSN)
  • /
  • 2288-5471(eISSN)

Domain

  • Nuclear Power ＞ Nuclear Fuel Cycle/Radioactive Waste Management

Aim & Scope

Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT) is a quarterly journal publishing peer-reviewed papers in various research areas associated with the nuclear fuel cycle and radioactive waste management. The journal is managed by the Korean Radioactive Waste Society. JNFCWT provides a forum for all stakeholders in the field of nuclear fuel cycle and radioactive waste management aiming at sharing and disseminating state of the art technical information, lessons learned from industrial experiences, and discussions on current socio-political issues. JNFCWT publishes the results of original research in the format of review articles, commentaries, and technical notes. From Volume 1 (2003) to Volume 11 (2013), this journal was published as the Journal of the Korean Radioactive Waste Society. It was renamed to the Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT) in 2013. As of 2019, the journal is listed in the SCOPUS of Elsevier. JNFCWT covers the following fields related to the nuclear fuel cycle and radioactive waste management as follows: 1. Policy, Safety, and Regulation of Nuclear Fuel Cycle / Non-proliferation, Safeguards, and Nuclear Security 2. Treatment, Transportation, and Storage of Spent Nuclear Fuel / Molten Salt Chemistry and Technology 3. High-Level Waste Immobilization and Disposal Technologies 4. Low- and Intermediate-Level Radioactive Waste Management 5. Decommissioning and Environmental Remediation 6. Radiation Protection, Shielding, and Environment 7. Radiochemistry and Nuclear Chemistry

http://www.krs.or.kr KSCI KCI SCOPUS

Volume 18 Issue spc

  1. Lee, Jongyoul;Kim, Inyoung;Ju, HeeJae;Choi, Heuijoo;Cho, Dongkeun 1
    https://doi.org/10.7733/jnfcwt.2020.18.S.1 PDF KSCI
    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%.
  2. Lee, Jongyoul;Kim, Hyeona;Kim, Inyoung;Choi, Heuijoo;Cho, Dongkeun 21
    https://doi.org/10.7733/jnfcwt.2020.18.S.21 PDF KSCI
    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.
  3. Kim, In-Young;Cho, Dong-Keun;Lee, Jongyoul;Choi, Heui-Joo 37
    https://doi.org/10.7733/jnfcwt.2020.18.S.37 PDF KSCI
    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.
  4. Choi, Heui-Joo;Lee, Jongyoul;Cho, Dongkeun 51
    https://doi.org/10.7733/jnfcwt.2020.18.S.51 PDF KSCI
    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.
  5. Jeong, Jongtae;Cho, Dong-Keun 63
    https://doi.org/10.7733/jnfcwt.2020.18.S.63 PDF KSCI
    The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS+ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS+ system as well as for the design modification and the operational safety assessment of the KRS+ system.
  6. Ju, Heejae;Kim, In-Young;Lee, Youn-Myoung;Kim, Jung-Woo;Hwang, Yongsoo;Choi, Heui-joo;Cho, Dong-Keun 75
    https://doi.org/10.7733/jnfcwt.2020.18.S.75 PDF KSCI
    The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS+), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS+ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10-4 mSv·yr-1, primarily from 129I. The proposed KRS+ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv·yr-1.