• Title/Summary/Keyword: nuclear fuel cycle

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Development of an Oxide Reduction Process for the Treatment of PWR Spent Fuel (PWR 사용후핵연료 처리를 위한 금속전환공정 개발)

  • Hur, Jin-Mok;Hong, Sun-Seok;Jeong, Sang-Mun;Lee, Han-Soo
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.8 no.1
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    • pp.77-84
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    • 2010

  • Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

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Nuclear Characteristics of a New(PWR-PHWR) Fuel Cycle (PWR-PHWR 핵연료 주기의 핵적 특성)

  • Jae Woong Song;Chang Hyun Chung
    • Nuclear Engineering and Technology
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    • v.17 no.3
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    • pp.185-192
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    • 1985

  • The fissile content of PWR spent fuel is higher than that of natural uranium which is normal fuel for CANDU type reactor. Investigated are the concepts of PWR spent fuel utilization in CANDU type reactor to diversify uranium resource and partially to solve storage problems of PWR spent fuel being gradually accumulated. Nuclear characteristics of uranium-plutonium mixed oxide fuel loaded in CANDU type reactor are analysed using the WIMS/D computer code. In this study, analyses are solely carried out upon the current CANDU type reactor design without changingany reactivity control devices.

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DEVELOPMENT OF THE ENIGMA FUEL PERFORMANCE CODE FOR WHOLE CORE ANALYSIS AND DRY STORAGE ASSESSMENTS

  • Rossiter, Glyn
    • Nuclear Engineering and Technology
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    • v.43 no.6
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    • pp.489-498
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    • 2011

  • UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  • https://doi.org/10.5516/NET.2011.43.6.489 인용 PDF KSCI