• Title, Summary, Keyword: DUPIC fuel

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A Study on the Micro-Focus X-Ray Inspection for Confirming the Soundness of End Closure Weld of DUPIC Fuel Elements (DUPIC 핵연료봉 봉단 용접부 건전성 확인을 위한 미세초점 X-선 투과시험에 관한 연구)

  • 김웅기;김수성;이정원;양명승
    • Journal of Welding and Joining
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    • v.19 no.1
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    • pp.88-94
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    • 2001
  • DUPIC (Direct use of spent PWR fuel in CANDU reactors) nuclear fuel is a CANDU fuel fabricated remotely from spent PWR fuel materials in a hot cell. The soundness of the end closure welds of nuclear fuel elements is an important factor for the safety and performance of nuclear fuel. To evaluate the soundness of the end closure welds of DUPIC fuel element, a precise X-ray inspection system is developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The fuel elements made of Zircaloy-4 and stainless steel by an Nd:YAG laser welding and a TIG welding aye inspected by the developed inspection system. The soundness of the welds of the fuel elements was confirmed by the X-ray inspection process, and the irradiation test of DUPIC fuel elements has been successfully completed at the HANARO research reactor.

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A Study on the Sintering of Simulated DUPIC Fuel (모의 DUPIC 핵연료의 소결 특성 연구)

  • 강권호;배기광;박희성;송기찬;문제선
    • Journal of Korean Powder Metallurgy Institute
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    • v.7 no.3
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    • pp.123-130
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    • 2000
  • The simulated DUPIC fuel provides a convenient way to investigate fuel properties and behaviours such as thermal conductivity, thermal expansion, fission gas release, leaching and so on. Several pellets simulating the composition and microstructure of the DUPIC fuel were fabricated from resintering powder through the OREOX process of the simulated spent fuel pellets, which were prepared from the mixture of stable forms of constituent nuclides. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent fuel was in agreement with the previous studies. The densities and the grain size of simulated DUPIC fuel was pellets are higher than those of simulated spent fuel pellets. Small metallic precipitates and oxide precipitates were observed on matrix grain boundaries.

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Study on Decay Characteristics Change of Spent Fuel Materials by DUPIC Fuel Cycle (DUPIC핵연료주기에 의한 사용 후 경수로핵연료의 방사선적 특성변화 분석)

  • Choi, Jong-Won;Ko, Won-Il;Lee, Jae-Sol;Park, Hyun-Soo
    • Journal of Radiation Protection and Research
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    • v.21 no.1
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    • pp.27-39
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    • 1996
  • The change in spent fuel characteristics by DUPIC fuel cycle(burnup of spent PWR fuel again in CANDU) is examined with time elapse since discharge. Major characteristics examined include isotopic concentration, radioactivity, decay heat radiotoxicity and radiation source-term of spent fuel material, which is existing in a type of spent PWR and DUPIC fuel. Behaviors of major nuclides contributing to such changes are also analyzed in terms of radionuclide concentration. From the analysis, the change in radionuclide concentration by DUPIC shows approximately 2% decrease in actinides concentration and 20% increase in fission products concentration. Radioactivity and decay heat of spent DUPIC fuel does not depend upon radionuclides concentrations, which is a unique in sence of general characteristics of spent fuel. In terms of gamma spectrum, spent DUPIC fuel shows lower values than that of spent PWR fuel by 40 to 50% in the range of $0.01{\sim}0.575$ MeV but much higher over 3.5MeV. Neutron Intensities of both spent fuels are mainly determined by $({\alpha},\;n)$ reaction and spontaneous fission reaction of actinides. Of them, especially, the spontaneous fission reaction Is a major neutron source-term, which causes that neutron intensities of spent DUPIC fuel $having{\sim}3.3$ times higher Cm-244 concentration are ${\sim}4$ times higher than that of spent PWR fuel.

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THE STATUS AND PROSPECT OF DUPIC FUEL TECHNOLOGY

  • Yang Myung-Seung;Choi Hang-Bok;Jeong Chang-Joon;Song Kee-Chan;Lee Jung-Won;Park Geun-Il;Kim Ho-Dong;Ko Won-Il;Park Jang-Jin;Kim Ki-Ho;Lee Ho-Hee;Park Joo-Hwan
    • Nuclear Engineering and Technology
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    • v.38 no.4
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    • pp.359-374
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    • 2006
  • Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

A SENSITIVITY STUDY ON NEUTRONIC PROPERTIES OF DUPIC FUEL

  • Park, Hangbok;Roh, Gyu-Hog
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.124-129
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    • 1998
  • A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

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Effect of DUPIC Cycle on CANDU Reactor Safety Parameters

  • Mohamed, Nader M.A.;Badawi, Alya
    • Nuclear Engineering and Technology
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    • v.48 no.5
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    • pp.1109-1119
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    • 2016
  • Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

Analysis of High Radioactive Materials in Irradiated DUPIC SIMFUEL Using EPMA (EPMA를 이용한 DUPIC 사용후 핵연료 핵분열 생성물의 특성 분석)

  • 정양홍;유병옥;주용선;이종원;정인하;김명한
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.2 no.2
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    • pp.125-133
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    • 2004
  • Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 ㎾/m of maximum linear power and 1,770 ㎿d/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 $\mu\textrm{m}$ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2∼2.5 $\mu\textrm{m}$ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.

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The Oxygen Potential of Urania Nuclear Fuel During Irradiation

  • Park, Kwang-Heon
    • The Korean Journal of Ceramics
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    • v.4 no.2
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    • pp.72-77
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    • 1998
  • A defect model for UO$_2$ fuel containing soluble fission products was devised based on the defect structure of pure and doped uranias. Using the equilibrium between fuel solid-solution and fission-products and the material balance within the fuel, a tracing method to get the stoichiometry change of urania fuel with burnup was made. This tracing method was applied to high burnup urania fuel and DUPIC fuel. The oxygen potential of urania fuel turned out to increase slightly with burnup. The stoichiometry change was calculated to be negligible due to the buffering role f Mo. The oxygen potential of DUPIC fuel out to be sensitive to the initial chemical state of Mo in the fuel.

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The Option Study of Oversea Shipment of DUPIC Fuel Elements to Canada (고방사성 산화물핵연료의 해외수송방안 분석)

  • 이호희;박장진;양명승;서기석
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.614-620
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    • 2003
  • KAERI has developed DUPIC nuclear fuel with the refabrication of spent PWR fuel discharged from domestic nuclear power plant by a dry process at M6 hot-cell in IMEF To verify the performance of DUPIC nuclear fuel, irradiation test at the operating conditions of commercial power plant is essential. Since the HANARO research reactor of KAERI does not have fuel test loop(FTL) for irradiating nuclear fuel under high temperature and high pressure conditions, DUPIC fuel cannot be irradiated in the FTL of HANARO. In the 13-th PRM among Korea, Canada, USA and IAEA, AECL proposed that KAERI fabricated DUPIC fuel can be irradiated in the FTL of the NRU research reactor without charge of neutrons. The transportation quantity of DUPIC fuel to Canada is 10 elements(about 6kg). This transportation package is classified as the 7-th class according to "recommendation on the transport of dangerous goods" made by the United Nations. In case of air shipment, until now, there is no proper air transportation cask for DUPIC fuel. In case of sea transportation is possible but requires very high cost.high cost.

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Robotic Floor Surface Decontamination System

  • Kim, Kiho;Park, Jangjin;Myungseung Yang
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • pp.133-134
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    • 2004
  • DUPIC (Direct Use of spent PWR fuel In CANDU) fuel cycle technology is being developed at Korea Atomic Energy Research Institute (KAERI). All the DUPIC fuel fabrication processes are remotely conducted in the completely shielded M6 hot-cell located in the Irradiated Material Examination Facility (IMEF) at KAERI. Undesirable products such as spent nuclear fuel powder debris and contaminated wastes are inevitably created during the DUPIC nuclear fuel fabrication processes.(omitted)

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