• Title/Summary/Keyword: Low-and intermediate-level radioactive waste

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A Study on the Shielding Analysis in Vitrification Facility of Low-and Intermediate Level Radioactive Wastes ($\cdot$저준위 방사성폐기물 유리화 시설의 차폐해석에 관한 연구)

  • 이창민;이건재;지평국;박종길;하종현;송명재
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2003.11a
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    • pp.524-531
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    • 2003
  • The usefulness of vitrification technology for low- and intermediate- level radioactive wastes was demonstrated because of high volume reduction, mechanical and chemical stability of final waste forms. Thus, a construction of the commercial vitrification plant Is currently promoted. Due to the high radiation level of the waste, shielding analysis has to be carried out for safe design in a vitrification facility. Because there has been no experience in the construction and operation of the vitrification facility in Korea, in this study, in order to get some information for help the detailed design and plan for operation in vitrification facility, shielding analysis for each facility in pilot plant is carried out by using source term from established study. For the selection of the shielding material, concrete was chosen compared to the lead because of economic advantage, weight consideration, and thermal resistance.

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Structural stability analysis of waste packages containing low- and intermediate-level radioactive waste in a silo-type repository

  • Byeon, Hyeongjin;Jeong, Gwan Yoon;Park, Jaeyeong
    • Nuclear Engineering and Technology
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    • v.53 no.5
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    • pp.1524-1533
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    • 2021
  • The structural stability of a waste package is essential for containing radioactive waste for the long term in a repository. A silo-type disposal facility would require more severe verification for the structural integrity, because of radioactive waste packages staked with several tens of meters and overburdens of crushed rocks and shotcretes. In this study, structural safety was analyzed for a silo-type repository, located approximately 100 m below sea level in Gyeongju, Korea. Finite element simulation was performed to investigate the influence of the loads from the backfilling materials and waste package stacks on the mechanical stress of the disposed of wastes and containers. It was identified that the current design of the waste package and the compressive strength criterion for the solidified waste would not be enough to maintain structural stability. Therefore, an enhanced criterion for the compressive strength of the solidified waste and several reinforced structural designs for the disposal concrete container were proposed to prevent failure of the waste package based on the results of parametric studies.

Radiological Impact Assessment for the Domestic On-road Transportation of Radioactive Isotope Wastes (방사성동위원소 폐기물의 국내육상운반에 관한 방사선영향 평가)

  • Seo, Myunghwan;Hong, Sung-Wook;Park, Jin Beak
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.14 no.3
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    • pp.279-287
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    • 2016
  • Korea Radioactive Waste Agency (KORAD) began to operate the low and intermediate level radioactive waste disposal facility in Gyeongju and to transport the radioactive waste containing radioactive isotopes from Daejeon to the disposal facility for the first time at 2015. For this radioactive waste transportation, in this study, radiological impact assessment is carried out for workers and public. The dose rate to workers and public during the transportation is estimated with consideration of the transportation scenarios and is compared with the Korean regulatory limit. The sensitivity analysis is carried out by considering both the variation of release ratios of the radioactive isotopes from the waste and the variation of the distances between the radioactive waste drum and worker during loading and unloading of radioactive waste. As for all the transportation scenarios, radiological impacts for workers and public have met the regulatory limits.

Simulation of the Migration of 3H and 14C Radionuclides on the 2nd Phase Facility at the Wolsong LILW Disposal Center

  • Ha, Jaechul;Son, Yuhwa;Cho, Chunhyung
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.18 no.4
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    • pp.439-455
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    • 2020
  • Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

Statistical Methodologies for Scaling Factor Implementation: Part 1. Overview of Current Scaling Factor Method for Radioactive Waste Characterization

  • Kim, Tae-Hyeong;Park, Junghwan;Lee, Jeongmook;Kim, Junhyuck;Kim, Jong-Yun;Lim, Sang Ho
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.18 no.4
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    • pp.517-536
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    • 2020
  • The radionuclide inventory in radioactive waste from nuclear power plants should be determined to secure the safety of final repositories. As an alternative to time-consuming, labor-intensive, and destructive radiochemical analysis, the indirect scaling factor (SF) method has been used to determine the concentrations of difficult-to-measure radionuclides. Despite its long history, the original SF methodology remains almost unchanged and now needs to be improved for advanced SF implementation. Intense public attention and interest have been strongly directed to the reliability of the procedures and data regarding repository safety since the first operation of the low- and intermediate-level radioactive waste disposal facility in Gyeongju, Korea. In this review, statistical methodologies for SF implementation are described and evaluated to achieve reasonable and advanced decision-making. The first part of this review begins with an overview of the current status of the scaling factor method and global experiences, including some specific statistical issues associated with SF implementation. In addition, this review aims to extend the applicability of SF to the characterization of large quantities of waste from the decommissioning of nuclear facilities.

Acceptance Criteria and Evaluation Techniques for Radioactive Waste Forms ( I ) (방사성폐기물 고화체의 인수기준 및 평가기술 ( I ))

  • 김정국;김준형;박헌휘
    • Nuclear Engineering and Technology
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    • v.23 no.1
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    • pp.81-94
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    • 1991
  • In order to develop the acceptance criteria for the low and intermediate level radioactive wastes for the land disposal: the following items were reviewed : classifications of radioactive wastes is respect to disposal, basic requirements and criteria that have to be considered during waste management from the origin to disposal. From these studies, the standard test methods to evaluate radioactive waste forms(or packages) were shown.

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Evaluation of cementation of intermediate level liquid waste produced from fission 99Mo production process and disposal feasibility of cement waste form

  • Shon, Jong-Sik;Lee, Hyun-Kyu;Kim, Tack-Jin;Kim, Gi-Yong;Jeon, Hongrae
    • Nuclear Engineering and Technology
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    • v.54 no.9
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    • pp.3235-3241
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    • 2022
  • The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR) for stable supply of 99Mo. The Fission 99Mo Production Process (FMPP) of KJRR produces solid waste such as spent uranium cake and alumina cake, and liquid waste in the form of intermediate level liquid waste (ILLW) and low level liquid waste (LLLW). This study thus established the operating range and optimum operating conditions for the cementation of ILLW from FMPP. It also evaluated whether cement waste form samples produced under optimum operational conditions satisfy the waste acceptance criteria (WAC) of a disposal facility in Korea (Korea radioactive waste agency, KORAD). Considering economic feasibility and safety, optimum operational conditions were achieved at a w/c ratio of 0.55, and the corresponding salt content was 5.71 wt%. The cement waste form samples prepared under optimum operational conditions were found to satisfy KORAD's WAC when tested for structural stability and leachability. The results indicate that the proposed cementation conditions for the disposal of ILLW from FMMP can be effectively applied to KJRR's disposal facility.

Service-life Prediction of Reinforced Concrete Structures in Subsurface Environment (지중 환경하에서의 철근콘크리트 구조물의 열화인자별 한계수명 평가)

  • Kwon, Ki-jung;Jung, Haeryong;Park, Joo-Wan
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.14 no.1
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    • pp.11-19
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    • 2016
  • This paper focuses on the estimation of durability and service-life of reinforced concrete structures in Wolsong Low- and intermediate-level wastes Disposal Center (WLDC) in Korea. There are six disposal silos located in the saturated environment. The silo concrete is degraded due to reactions with groundwater and chemical attacks, and finally it will lose its properties as a transport barrier. The infiltration of sulfate and magnesium, leaching of potassium hydroxide, and chlorine induced corrosion are the most significant factors for degradation of reinforced concrete structure in underground environment. From the result of evaluation of the degradation time for each factor, the degradation rate of the reinforced concrete due to sulfate and magnesium is $1.308{\times}10^{-3}cm/yr$, and it is estimated to take 48,000 years for full degradation while potassium hydroxide is leached in depth of less than 1.5 cm at 1,000 years after the initiation of degradation. In case of chlorine induced corrosion, it takes 1,648 years to initiate corrosion in the main reinforced bar and 2,288 years to reach the lifetime limit of the structural integrity, and thus it is evaluated as the most significant factor.

Concrete Degradation Comparison of Computer Programs for Post-Closure Safety Assessment of Wolsong Low-and Intermediate-Level Radioactive Waste Disposal Facility (월성원자력환경관리센터 폐쇄 후 안전평가 컴퓨터프로그램의 콘크리트 열화현상에 대한 상호비교)

  • Jung, Kang-Il;Bang, Je-Heon;Park, Jin Beak;Yoon, Jeong Hyoun
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.11 no.4
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    • pp.311-324
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    • 2013
  • To ensure the reliability of computer programs used for the post-closure safety assessment in the Wolsong LILW Center, the results from MASCOT, SAFE-ROCK and GOLDSIM programs are compared with a problem for degradation. Advantages and disadvantages of each computer programs are individually analyzed. Effects on the individual dose are assessed with each computer programs. MASCOT and SAFE-ROCK showed similar results for $^{129}I$ and $^3H$. However, GOLDSIM represented different results for $^{129}I$ and $^3H$. It is analyzed further and compared with the fluxes in each barrier of the disposal system. Througout the benchmarking testing of the computer program, the limitation of computer program can be continuously found out for the mature post-closure safety of Korean radwaste disposal system.

A Safety Assessment for the Wolsong LILW Disposal Center: As a part of safety case for the first stage disposal (월성원자력환경관리센터의 폐쇄후 처분안전성평가: 1단계 인허가 적용사례를 중심으로)

  • Park, Joo-Wan;Yoon, Jeong-Hyun;Kim, Chang-Lak
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.6 no.4
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    • pp.329-346
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    • 2008
  • Post-closure safety assessment for the Wolsong Low- and Intermediate-level radioactive waste Disposal Center is described. Based on assessment context, closure concept and ground water flow characteristics of the disposal site, brief descriptions are included on the assessment scenarios, models, input parameters and tools. Radionuclide transport modeling in the near-field and far-field, gas generation and transport modeling, human intrusion and biosphere transport are also described briefly. Assessment results for each scenarios are shown to meet the performance criteria of regulatory body. Further and continuous efforts to improve the safety of disposal facility will be made during the construction and operational period.

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