• Title, Summary, Keyword: Dry interim storage facility

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Development of a Probabilistic Safety Assessment Framework for an Interim Dry Storage Facility Subjected to an Aircraft Crash Using Best-Estimate Structural Analysis

  • Almomani, Belal;Jang, Dongchan;Lee, Sanghoon;Kang, Hyun Gook
    • Nuclear Engineering and Technology
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    • v.49 no.2
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    • pp.411-425
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    • 2017
  • Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose-risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

Technology Trends in Spent Nuclear Fuel Cask and Dry Storage (사용후핵연료 운반용기 및 건식저장 기술 동향)

  • Shin, Jung Cheol;Yang, Jong Dae;Sung, Un Hak;Ryu, Sung Woo;Park, Yeong Woo
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.16 no.1
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    • pp.110-116
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    • 2020
  • As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

Preliminary Analysis of Dose Rate Variation on the Containment Building Wall of Dry Interim Storage Facilities for PWR Spent Nuclear Fuel (경수로 사용후핵연료 건식 중간저장시설의 격납건물 크기에 따른 건물 벽면에서의 방사선량률 추이 예비 분석)

  • Seo, M.H.;Yoon, J.H.;Cha, G.Y.
    • Journal of Radiation Protection and Research
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    • v.38 no.4
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    • pp.189-193
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    • 2013
  • Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

Preliminary Assessment of Radiation Impact from Dry Storage Facilities for PWR Spent Fuel (경수로 사용후핵연료 건식 중간저장시설에 대한 예비 방사선 영향 평가)

  • Kim, T.M.;Baeg, C.Y.;Cha, G.Y.;Lee, W.G.;Kim, S.Y.
    • Journal of Radiation Protection and Research
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    • v.37 no.4
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    • pp.197-201
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    • 2012
  • Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

Development for Improvement Methodology of Radiation Shielding Evaluation Efficiency about PWR SNF Interim Storage Facility (PWR 사용후핵연료 중간저장시설의 몬테칼로 차폐해석 방법에 대한 계산효율성 개선방안 연구)

  • Kim, Taeman;Seo, Myungwhan;Cho, Chunhyung;Cha, Gilyong;Kim, Soonyoung
    • Journal of Radiation Protection and Research
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    • v.40 no.2
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    • pp.92-100
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    • 2015
  • For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

Preliminary Shielding Analysis of the Concrete Cask for Spent Nuclear Fuel Under Dry Storage Conditions (건식저장조건의 사용후핵연료 콘크리트 저장용기 예비 방사선 차폐 평가)

  • Kim, Tae-Man;Dho, Ho-Seog;Cho, Chun-Hyung;Ko, Jae-Hun
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.15 no.4
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    • pp.391-402
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    • 2017
  • The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

Technology for AR Dry Storage of Spent Fuel (원전부지내 사용후핵연료 건식저장기술 분석)

  • Lee, Heung-Young;Yoon, Suk-Jung;Lee, Ik-Hwan;Seo, Ki-Seog
    • Journal of Radiation Protection and Research
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    • v.21 no.4
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    • pp.313-327
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    • 1996
  • As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

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