• Title, Summary, Keyword: Spent nuclear fuel (SNF)

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Innovative technologies for spent fuel safe management at Ignalina channel-type reactors

  • Babilas, Egidijus;Dokucajev, Pavel;Janulevicius, Darius;Markelov, Aleksej;Pabarcius, Raimondas;Rimkevicius, Sigitas;Uspuras, Eugenijus;Vaisnoras, Mindaugas
    • Nuclear Engineering and Technology
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    • v.50 no.3
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    • pp.504-511
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    • 2018
  • In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

Requirements for the Transportation of Spent Nuclear Fuel (SNF) in Terms of Fuel Integrity and Data Needed According to

  • Noh, J.S.;Kim, Y.K.;Kim, T.W.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.115-116
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    • 2017
  • For the safe transportation of SNF and licensing, the integrity of SNF should be evaluated carefully. Researches to obtain the data for SNF cladding properties, i.e. impact toughness, DBTT (hydride behavior) when evaluating transportation of SNF, shall be precisely implemented by simulating the condition of real SNF to the hilt, accordingly.

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Calculation of the Radioactivity of Spent Nuclear Fuel by Using a Spline Interpolation (스플라인 보간법을 이용한 사용후핵연료의 방사능 계산)

  • Jin, Sungwan;Kim, Yong-Uhn
    • New Physics: Sae Mulli
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    • v.67 no.8
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    • pp.947-951
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    • 2017
  • The radioactivity of a chain-decaying source such as some radioactive isotopes contained in spent nuclear fuel (SNF) can be calculated either by solving the Bateman equation, by calculating the matrix exponential function, by using a Monte Carlo computation method, or by using a numerical integration method. When these methods are applied, the computation time exponentially increases as the number of generations of the damping chain increases. In this study, the radioactivity of SNF as a function time was obtained by using a spline interpolation method that is used to express vector images in computer graphics. We measured the computation time by increasing the number of data points and calculating the frequency of 'Not a Number error' in the calculation results obtained by using the five different methods. When the spline method was applied, the radioactivity of SNF as a function of time could be obtained in a shorter time compared to the other methods.

Korean Status and Prospects for Radioactive Waste Management

  • Song, M.J.
    • Journal of Nuclear Fuel Cycle and Waste Technology
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    • v.1 no.1
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    • pp.1-7
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    • 2013
  • The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. Since the initial introduction of nuclear power to Korea in 1978, rapid growth in nuclear power has been achieved. This large nuclear power generation program has produced a significant amount of radioactive waste, both low- and intermediate-level waste (LILW) and spent nuclear fuel (SNF); and the amount of waste is steadily growing. For the management of LILW, the Wolsong LILW Disposal Center, which has a final waste disposal capacity of 800,000 drums, is under construction, and is expected to be completed by June 2014. Korean policy about how to manage the SNF has not yet been decided. In 2004, the Atomic Energy Commission decided that a national policy for SNF management should be established considering both technological development and public consensus. Currently, SNF is being stored at reactor sites under the responsibility of plant operator. The at-reactor SNF storage capacity will run out starting in 2024. In this paper, the fundamental principles and steps for implementation of a Korean policy for national radioactive waste management are introduced. Korean practices and prospects regarding radioactive waste management are also summarized, with a focus on strategy for policy-making on SNF management.

Dry storage of spent nuclear fuel and high active waste in Germany-Current situation and technical aspects on inventories integrity for a prolonged storage time

  • Spykman, Gerold
    • Nuclear Engineering and Technology
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    • v.50 no.2
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    • pp.313-317
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    • 2018
  • Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

Projection and Burnup Trends of Spent Nuclear Fuel in Korea (국내 사용후핵연료 현황 분석)

  • 조동건;최종원;이희환
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.261-267
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    • 2004
  • Inventories, projections, and characteristics of spent nuclear fuel(SNF) generated from domestic nuclear power plants were updated to support high-level waste disposal system design. The historical and projected inventory by the end 2055 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively The ratio of quantity for TEX>$17{\times}17$ SNF was shown to be 0.6 as of 2003. The amount of TEX>$17{\times}17$ SNF, however, will be less than that of TEX>$16{\times}16$ KSFA after 2012, while the quantity of TEX>$16{\times}16$ KSFA will reach to 70% of the total spent fuels in the 2055. Average turnup of SNF revealed ~36GWD/MTU and ~40GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will exceed 45GWD/MTU at the end of 2000's. Therefore, it seems reasonable to use the TEX>$17{\times}17$ 4.5w/o, 45GWD/MTU as the Reference SNF at present state. The TEX>$16{\times}16$ KSFA 4.5w/o, 55GWD/MTU, however, should be Reference SNF after ~2010.

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Spent Nuclear Fuel Safety Evaluation Methodology (SSEM) for Storage and Transportation

  • Kim, Y.K.;Noh, J. S.;Lee, S.K.;Kim, T.W.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.57-58
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    • 2017
  • New approach to achieve the safety goals in transportation and dry storage of SNF, so called SSEM has been proposed. The main concept of the SSEM is that it simplifies the reviewing processes of each campaign of the transportation or storage of SNF with standard format. This SSEM could be considered as a model case for assuring public that the SNF be managed safely.

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Parametric study on the structural response of a high burnup spent nuclear fuel rod under drop impact considering post-irradiated fuel conditions

  • Almomani, Belal;Kim, Seyeon;Jang, Dongchan;Lee, Sanghoon
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.1079-1092
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    • 2020
  • A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

Reference Spent Nuclear Fuel for Pyroprocessing Facility Design (파이로공정 시설 개념설계를 위한 기준 사용후핵연료 선정)

  • Cho, Dong-Keun;Yoon, Seok-Kyun;Choi, Heui-Joo;Choi, Jong-Won;Ko, Won-Il
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.6 no.3
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    • pp.225-232
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    • 2008
  • An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight, $^{235}U$ enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial $^{235}U$ enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16$\times$16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16$\times$16 Korean Standard Fuel Assembly with a cross section of 21.4 cm$\times$21.4 cm, a length of 453 cm, a mass of 672 kg, the initial $^{235}U$ enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU.

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Validation of nuclide depletion capabilities in Monte Carlo code MCS

  • Ebiwonjumi, Bamidele;Lee, Hyunsuk;Kim, Wonkyeong;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.52 no.9
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    • pp.1907-1916
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    • 2020
  • In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.