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

2008 ◽  
Author(s):  
R. Calabrese ◽  
F. Vettraino ◽  
T. Tverberg
Keyword(s):  
Thermal Conductivity ◽  
Fuel Cycle ◽  
Data Set ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Rod Bundle ◽  
Pile Testing ◽  
Inert Matrix Fuels ◽  
Irradiation Performance ◽  
Semi Empirical

Inert matrix fuels are a possible option to reduce separated plutonium stockpiles by burning it in LWR fleet. A high burning efficiency targeted by preventing new plutonium build-up 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-stabilised 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-stabilised zirconia under-irradiation performance is still to be fully assessed, with this regard the thermal conductivity, markedly lower than UOX and 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, that have been performed in the OECD Halden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90–97% of the 45 MWd/kgUeq 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 matrix, the fissile phase being HEU oxide (UO2 93% 235U enriched). IMT is a ternary fuel composed by csz+thoria matrix and HEU oxide as 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 behaviour by means of the TRANSURANUS code. Thermal conductivity and its degradation with burnup, densification-swelling response and FGR are tentatively modelled in the burnup domain of IFA-652. In particular it is pointed out the effects of pellet geometry and fuel microstructures in the IM and IMT cases. 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 modelling of the IMF is still a work in progress.

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

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
R. Calabrese ◽  
F. Vettraino ◽  
T. Tverberg
Keyword(s):  
Thermal Conductivity ◽  
Fuel Cycle ◽  
Stabilized Zirconia ◽  
Data Set ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Rod Bundle ◽  
Pile Testing ◽  
Enriched Uranium ◽  
Irradiation Performance

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 45MWd∕kgUeq 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.

Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

2006 ◽  
Author(s):  
R. Calabrese ◽  
F. Vettraino ◽  
T. Tverberg
Keyword(s):  
Experimental Data ◽  
Thermal Conductivity ◽  
Early Stage ◽  
Fuel Temperature ◽  
Swelling Behaviour ◽  
Pressure Transducers ◽  
Inert Matrix ◽  
Pile Testing ◽  
Inert Matrix Fuels ◽  
Irradiation Cycle

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd·kgUeq−1 vs. 45 MWd·kgUeq−1 (40 MWd·kgUOXeq−1) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd·kgUeq−1) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed.

Reprocessing Silicon Carbide Inert Matrix Fuel by Molten Salt Corrosion

2009 ◽  
Vol 1215 ◽  
Author(s):  
Ting Cheng ◽  
Ronald Baney ◽  
James Tulenko
Keyword(s):  
Silicon Carbide ◽  
Matrix Material ◽  
Water Soluble ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Molten Carbonates ◽  
Air Atmosphere ◽  
Salt Corrosion ◽  
Bulk Reaction ◽  
Inert Matrix Fuels

AbstractSilicon carbide is one of the prime matrix material candidates for inert matrix fuels (IMF) which are being designed to reduce plutonium and long half-life actinide inventories through transmutation. Since complete transmutation is impractical in a single in-core run, reprocessing the inert matrix fuels becomes necessary. The current reprocessing techniques of many inert matrix materials involve dissolution of spent fuels in acidic aqueous solutions. However, SiC cannot be dissolved by that process. Thus, new reprocessing techniques are required.This paper discusses a possible way for separating transuranic (actinide) species from a bulk silicon carbide (SiC) matrix utilizing molten carbonates. Bulk reaction-bonded SiC and SiC powder (1 μm) were corroded at high temperatures (above 850 °C) in molten carbonates (K2CO3 and Na2CO3) in an air atmosphere to form water soluble silicates. Separation of Ceria (used as a surrogate for the plutonium fissile fuel) was achieved by dissolving the silicates in boiling water and leaving behind the solid ceria (CeO2).

Inert matrix fuel — A new challenge for material technology in the nuclear fuel cycle

2001 ◽  
Vol 7 (2) ◽  
pp. 159-164 ◽  
Author(s):  
Young-Woo Lee ◽  
Chang Young Joung ◽  
Si Hyung Kim ◽  
Sang-Chul Lee
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Material Technology ◽  
Inert Matrix Fuel ◽  
Inert Matrix
Sign in / Sign up

Export Citation Format

Share Document