• 제목/요약/키워드: fuel burnup

검색결과 260건 처리시간 0.027초

Experimental validation of a nuclear forensics methodology for source reactor-type discrimination of chemically separated plutonium

  • Osborn, Jeremy M.;Glennon, Kevin J.;Kitcher, Evans D.;Burns, Jonathan D.;Folden, Charles M. III;Chirayath, Sunil S.
    • Nuclear Engineering and Technology
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    • 제51권2호
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    • pp.384-393
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    • 2019

  • An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed.

  • https://doi.org/10.1016/j.net.2018.11.003 인용 PDF KSCI

RESULTS OF THERMAL CREEP TEST ON HIGHLY IRRADIATED ZIRLO

  • Quecedo, M.;Lloret, M.;Conde, J.M.;Alejano, C.;Gago, J.A.;Fernandez, F.J.
    • Nuclear Engineering and Technology
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    • 제41권2호
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    • pp.179-186
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    • 2009

  • This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

  • https://doi.org/10.5516/NET.2009.41.2.179 인용 PDF KSCI

HIGH BURNUP FUEL TECHNOLOGY IN KOREA

  • Song, Kun-Woo;Jeong, Yong-Hwan;Kim, Keon-Sik;Bang, Je-Geon;Chun, Tae-Hyun;Kim, Hyung-Kyu;Song, Kee-Nam
    • Nuclear Engineering and Technology
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    • 제40권1호
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    • pp.21-36
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    • 2008

  • High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

  • https://doi.org/10.5516/NET.2008.40.1.021 인용 PDF KSCI