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Research Papers: Nuclear Power

Inert Matrix Fuel Analysis by Means of the TRANSURANUS Code: The Halden IFA-652 In-Pile Test

[+] Author and Article Information
R. Calabrese

ENEA, Innovative Nuclear Systems and Fuel Cycle Closure Division, via Martiri di Monte Sole, 4, 40129 Bologna, Italyrolando.calabrese@bologna.enea.it

F. Vettraino

ENEA, Innovative Nuclear Systems and Fuel Cycle Closure Division, via Martiri di Monte Sole, 4, 40129 Bologna, Italy

T. Tverberg

 OECD Halden Reactor Project, P.O. Box 173, 1751 Halden, Norway

J. Eng. Gas Turbines Power 131(1), 012907 (Oct 02, 2008) (10 pages) doi:10.1115/1.2983140 History: Received July 25, 2008; Revised August 05, 2008; Published October 02, 2008

Inert matrix fuels (IMFs) are a possible option to reduce separated plutonium stockpiles by burning it in light water reactor (LWR) fleet. A high burning efficiency targeted by preventing new plutonium buildup under irradiation (U-free fuel), a proved high radiation damage, and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Among other options, both calcia-stabilized zirconia (CSZ) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. While several in-pile tests concerning thoria fuels are found, calcia-stabilized zirconia under-irradiation performance is still to be fully assessed; with this regard the thermal conductivity, markedly lower than the uranium oxide (UOX) and mixed oxide (MOX) cases, plays a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing of three different U-free inert matrix fuel concepts, which have been performed in the OECD Halden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90–97% of the 45MWdkgUeq target under typical LWR irradiation conditions. The test rig is a six-rod bundle loaded with IM, IMT, and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria as matrices, the fissile phase being the high enriched uranium (HEU) oxide (UO2 93% U235 enriched). IMT is a ternary fuel composed by CSZ+thoria matrix and HEU oxide as a fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centerline temperature, pin inner pressure, and fuel stack elongation measurements. Our purpose is to investigate the key processes of IMF under-irradiation behavior by means of the TRANSURANUS fuel performance code. Thermal conductivity and its degradation with burnup, densification-swelling response, and fission gas release (FGR) are tentatively modeled in the burnup range of IFA-652. In particular, the effects of pellet geometry and fuel microstructures in the IM and IMT cases are pointed out. The consistency of our results is discussed aiming at understanding the in-pile response, as a fundamental step, in the perspective of future deployment of the nuclear fuels we are dealing with. Notwithstanding this ambitious objective, it is clear, however, that these results rely on a limited data set and that, as TRANSURANUS is a semi-empirical code mostly tailored for commercial fuels, the modeling of the IMF is still a work in progress.

Copyright © 2009 by American Society of Mechanical Engineers
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References

Figures

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Figure 1

IFA-652 test rig (cross section)

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Figure 2

Average linear heat rate and power of the IFA-652 rig

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Figure 3

IMF energy dispersive-X surface mapping

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Figure 4

Fully dense thermal conductivity (8,12)

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Figure 5

Fuel centerline temperature—rod 1 (±10% error band)

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Figure 6

Fuel centerline temperature—rod 2 (±10% error band)

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Figure 7

Fuel centerline temperature—rod 6 (±10% error band)

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Figure 8

Ratio of predicted fuel centerline temperature and inner pressure to experimental temperature and inner pressure—rod 1 (±10% error band)

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Figure 9

Ratio of predicted fuel centerline temperature and inner pressure to experimental temperature and inner pressure—rod 2 (±10% error band)

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Figure 10

Ratio of predicted fuel centerline temperature and inner pressure to experimental temperature and inner pressure—rod 6 (±10% error band)

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Figure 11

HSB inner pressure—rod 1

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Figure 12

HSB inner pressure—rod 2

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Figure 13

HSB inner pressure—rod 6

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Figure 14

HSB fuel stack elongation—rod 2

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Figure 15

HSB fuel stack elongation—rod 6

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Figure 17

Predicted HSB gap width (slice 5)

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