• Title/Summary/Keyword: Spent fuel rod

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A Study on the Local Boiling of the Consolidated Spent Fuel Storage Pool (조밀화된 사용후 핵연료 저장조에서의 국부 비등에 관한 연구)

  • Lee, Chang-Ju;Lee, Kun-Jai
    • Nuclear Engineering and Technology
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    • v.25 no.1
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    • pp.8-19
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    • 1993
  • The natural convection model of the consolidated system has been developed to make sure the removal of decay heat generated in the spent fuel for the loss of forced cooling accident. The numerical technique employed was based on the ADI scheme. The calculation of heat generation rate in the spent fuel was peformed by the ANS-79 decay heat model, and the nonuniform surface heat flux is assumed with a chopped sine curve for the conservative decay heat generation input. The sensitivity study was performed to examine the possibility of the pool bulk boiling by varying the various parameters, i.e. inter-fuel spacing ratio, heat generation power, and radius of the fuel rod. The application results of this model show that the natural circulation flow through compacted spent fuel bundles enables the pool temperature to control in a safe and effective manner, after the required cooling time. The corresponding acceptance criteria of the cooling time for rearranging the spent fuel rods were also found.

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Thermal Analyses of Deep Geological Disposal Cell With Heterogeneous Modeling of PLUS7 Spent Nuclear Fuel

  • Hyungju Yun;Min-Seok Kim;Manho Han;Seo-Yeon Cho
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.21 no.4
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    • pp.517-529
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    • 2023
  • The objectives of this paper are: (1) to conduct the thermal analyses of the disposal cell using COMSOL Multiphysics; (2) to determine whether the design of the disposal cell satisfies the thermal design requirement; and (3) to evaluate the effect of design modifications on the temperature of the disposal cell. Specifically, the analysis incorporated a heterogeneous model of 236 fuel rod heat sources of spent nuclear fuel (SNF) to improve the reality of the modeling. In the reference case, the design, featuring 8 m between deposition holes and 30 m between deposition tunnels for 40 years of the SNF cooling time, did not meet the design requirement. For the first modified case, the designs with 9 m and 10 m between the deposition holes for the cooling time of 40 years and five spacings for 50 and 60 years were found to meet the requirement. For the second modified case, the designs with 35 m and 40 m between the deposition tunnels for 40 years, 25 m to 40 m for 50 years and five spacings for 60 years also met the requirement. This study contributes to the advancement of the thermal analysis technique of a disposal cell.

Analysis of Characteristics of Spent Fuels on Long-Term Dry Storage Condition

  • Yoon, Suji;Park, Kwangheon;Yun, Hyungju
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.19 no.2
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    • pp.205-214
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    • 2021
  • Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

Preliminary study of artificial intelligence-based fuel-rod pattern analysis of low-quality tomographic image of fuel assembly

  • Seong, Saerom;Choi, Sehwan;Ahn, Jae Joon;Choi, Hyung-joo;Chung, Yong Hyun;You, Sei Hwan;Yeom, Yeon Soo;Choi, Hyun Joon;Min, Chul Hee
    • Nuclear Engineering and Technology
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    • v.54 no.10
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    • pp.3943-3948
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    • 2022
  • Single-photon emission computed tomography is one of the reliable pin-by-pin verification techniques for spent-fuel assemblies. One of the challenges with this technique is to increase the total fuel assembly verification speed while maintaining high verification accuracy. The aim of the present study, therefore, was to develop an artificial intelligence (AI) algorithm-based tomographic image analysis technique for partial-defect verification of fuel assemblies. With the Monte Carlo (MC) simulation technique, a tomographic image dataset consisting of 511 fuel-rod patterns of a 3 × 3 fuel assembly was generated, and with these images, the VGG16, GoogLeNet, and ResNet models were trained. According to an evaluation of these models for different training dataset sizes, the ResNet model showed 100% pattern estimation accuracy. And, based on the different tomographic image qualities, all of the models showed almost 100% pattern estimation accuracy, even for low-quality images with unrecognizable fuel patterns. This study verified that an AI model can be effectively employed for accurate and fast partial-defect verification of fuel assemblies.

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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Preliminary data analysis of surrogate fuel-loaded road transportation tests under normal conditions of transport

  • JaeHoon Lim;Woo-seok Choi
    • Nuclear Engineering and Technology
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    • v.54 no.11
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    • pp.4030-4048
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    • 2022
  • In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

Thermal Analysis of a Spent Fuel Storage Cask under Normal and Off-Normal Conditions

  • Lee, J. C.;K. S. Bang;K. S. Seo;Kim, H.D.;Park, B. I.;Lee, H. Y.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2003.11a
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    • pp.601-608
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    • 2003
  • Thermal analyses have been carried out for a spent fuel dry storage cask under normal and off-normal conditions. Environmental temperature is assumed to be $15^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of $38^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition. Temperature distributions for the off-normal conditions were slightly higher than the normal conditions.

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Generation of Group Constant of Fission Product for Criticality Analysis of Spent Fuel (사용후 핵연료의 핵임계도 분석에 필요한 핵분열생성물의 핵군단면적 생산)

  • Shin, H.S.;Choi, B.I;Park, J.M.;Ro, S.G.
    • Journal of Radiation Protection and Research
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    • v.14 no.2
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    • pp.30-36
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    • 1989
  • A FISSLIB, 51 group nuclear data set for 22 product nuclides, which are present in spent fuel and significantly affect the criticality of spent fuel, was generated from ENDF/B-IV using XLACS-II. The FISSLIB is ready to be used together with a data set generated from DLC-43/CSRL using AMPX system. The reliability of FISSLIB was verified by comparison with the data reported in BNL-325. Using FISSLIB, the criticality of KORI-1 spent fuel rod arranged infinitely was analyzed, and it was found that $K_{eff}$ of the spent fuel including fission products was lower by 9-14% than that calculated without fission products.

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Thermal Analysis on the Spent Fuel Shipping Cask for a PWR Fuel Assembly (PWR 사용후 핵연료 수송용기에 대한 열해석)

  • Hee Yung Kang;Eun Ho Kwack;Byung Jin Son
    • Nuclear Engineering and Technology
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    • v.15 no.4
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    • pp.248-255
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    • 1983
  • The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

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