Feasibility study of a moderation-enhanced reactor core loaded 100% with MOX fuel감속이 증진된 혼합산화물 연료 100% 장전 원자로 노심의 타당성 연구

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The mixed-oxide (MOX) fuel assemblies have been partially loaded and operated successfully in some commercial pressurized water reactors (PWRs) and boiling water reactors (BWRs). Lately, for the more efficient utilization of nuclear fuel resources and, in particular, for the disposition of weapons-grade plutonium, there has been a growing interest in full loading of MOX fuels in existing reactors (if required) with only minor modification. In this study, a PWR core loaded 100% with moderation-enhanced mixed-oxide fuel (hereafter called "meMOX" fuel in contrast to the standard mixed-oxide fuel designated here as "stMOX" or simply "MOX" fuel) is studied. To alleviate the spectrum hardening due to the larger capture-to-fission ratio of plutonium, the meMOX fuel assembly is obtained from the stMOX fuel assembly by removing several fuel rods (e.g., 36 fuel rods in the $17 \times 17$ fuel assembly are replaced by water holes). This increases the moderator-to-fuel volume ratio from 2.02 to 2.51, thus enhancing moderation of the neutrons. This study also includes the determination of equivalent plutonium content in meMOX or stMOX fuel, that is equivalent in cycle or discharge burnup with a prescribed conventional $UO_2$ fuel. The isotopic composition of the plutonium in twelve different cases used to obtain the equivalent plutonium content ranges from reactor-grade plutonium with about 55- 85 w/o of fissile plutonium to weapons-grade plutonium with about 94 w/o of fissile plutonium. Both the meMOX assembly characteristics and the core (meMOX loaded Ulchin Unit 1) characteristics are provided. The assemblywise burnup-dependent neutronic characteristics consist of assembly reactivity, inventories of major isotopes, control rod worth, moderator temperature coefficient (MTC), Doppler temperature coefficient (DTC), differential boron worth, void coefficient of reactivity, and peak rod power within an assembly, etc. These were obtained by the assembly depletion code CASMO-3....
Cho, Nam-Zinresearcher조남진researcher
한국과학기술원 : 원자력공학과,
Issue Date
105544/325007 / 000943093

학위논문(석사) - 한국과학기술원 : 원자력공학과, 1996.2, [ iv, 46 p. ]


Moderation-Enhanced Reactor Core; MOX Fuel

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