• Title/Summary/Keyword: MOX

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Benchmark Calculations of Lattice Codes for the Doppler Coefficient of MOX Fuel

  • Shin, Ho-Cheol;Bae, Sung-Man;Kim, Yong-Bae;Lee, Sang-Hee
    • Proceedings of the Korean Nuclear Society Conference
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    • 1996.05a
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    • pp.46-51
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    • 1996

  • In this study we calculate the infite multiplication factors ($k_{\infty}$) and the Doppler temperature coefficients (DTC) of two mixed-oxide (MOX) fuel rods with different plutonium contents by using PHOENIX-P, HELIOS and CASMO-3 codes. The results were compared against the reference values obtained by MCNP-3A continuous-energy Monte Carlo code. The purpose of this study is to benchmark the accuracy of these lattice codes. The PHOENIX-P's Doppler coefficients calculated were in good agreement with the MCNP results within the Monte-Carlo uncertainty band which is in the order of $\pm$ 10% for the Doppler coefficients..

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Transient analysis of a subcritical reactor core with a MOX-Fuel using the birth-and-death model

  • Korbu, Tamara;Kuzmin, Andrei;Rudak, Eduard;Kravchenko, Maksim
    • Nuclear Engineering and Technology
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    • v.53 no.6
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    • pp.1731-1735
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    • 2021

  • The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for 239Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

  • https://doi.org/10.1016/j.net.2020.11.023 인용 PDF KSCI

PROGRESS IN NUCLEAR FUEL TECHNOLOGY IN KOREA

  • Song, Kun-Woo;Jeon, Kyeong-Lak;Jang, Young-Ki;Park, Joo-Hwan;Koo, Yang-Hyun
    • Nuclear Engineering and Technology
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    • v.41 no.4
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    • pp.493-520
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    • 2009

  • During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

  • https://doi.org/10.5516/NET.2009.41.4.493 인용 PDF KSCI

해외 정보 - 일본의 플루토늄 문제를 해결하기 위한 현실적 접근

  • Acton, James M.
    • Nuclear industry
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    • v.36 no.1
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    • pp.75-76
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    • 2016

  • 일본은 사용후연료의 재처리 정책을 진행하고 있으나 고속증식로 계획이 지연됨에 따라 그 처리의 목표가 확실하지 못하다. MOX(Mixed-Oxide Fuel : 혼합 산화물 연료)에 의한 처리가 계획되어 있으나 상당히 어려운 상황이다. 또 2018년에 미 일 원자력협력협정의 갱신이 예정되어 있는데 이 플루토늄의 취급이 문제가 될 예정이다. 미국은 핵무기의 원료나 테러에 사용할지도 모르는 플루토늄의 확산을 경계하고 있기 때문이다.

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