scholarly journals Reducing Actinide Production Using Inert Matrix Fuels

2017 ◽  
Author(s):  
Mark Deinert
Keyword(s):  
Inert Matrix ◽  
Inert Matrix Fuels

Fabrication and characterization of silicon nitride-based inert matrix fuels sintered with magnesium silicates

2011 ◽  
Vol 53 (7) ◽  
pp. 1078-1081
Author(s):  
Toshiyuki Usuki ◽  
Katsumi Yoshida ◽  
Toyohiko Yano ◽  
Shuhei Miwa ◽  
Masahiko Osaka
Keyword(s):  
Silicon Nitride ◽  
Inert Matrix ◽  
Magnesium Silicates ◽  
Fabrication And Characterization ◽  
Inert Matrix Fuels

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).

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.

Mechanical behaviour of macro-dispersed inert matrix fuels

2003 ◽  
Vol 317 (2-3) ◽  
pp. 217-225 ◽  
Author(s):  
E.A.C. Neeft ◽  
K. Bakker ◽  
R.L. Belvroy ◽  
W.J. Tams ◽  
R.P.C. Schram ◽  
...  
Keyword(s):  
Mechanical Behaviour ◽  
Inert Matrix ◽  
Inert Matrix Fuels

The fabrication and irradiation of plutonium-containing inert matrix fuels for the ‘Once Though Then Out’ experiment

2003 ◽  
Vol 319 ◽  
pp. 118-125 ◽  
Author(s):  
R.P.C Schram ◽  
R.R van der Laan ◽  
F.C Klaassen ◽  
K Bakker ◽  
T Yamashita ◽  
...  
Keyword(s):  
Inert Matrix ◽  
Inert Matrix Fuels

Inert Matrix Fuels

2001 ◽  
pp. 4066-4068
Author(s):  
N. Chauvin ◽  
M. Pelletier
Keyword(s):  
Inert Matrix ◽  
Inert Matrix Fuels
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