scholarly journals Progress on Synthesis of Low Content Inert Matrix Fuel Pellets

2019 ◽  
Author(s):  
Rachel Seibert ◽  
Andrew Nelson
Keyword(s):  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

On the basic properties of an iron-based simulated cermet inert matrix fuel, synthesized by a dry route in oxidizing conditions

2006 ◽  
Vol 48 (6) ◽  
pp. 590-598 ◽  
Author(s):  
N. Kamel ◽  
H. Aït-Amar ◽  
Z. Kamel ◽  
N. Souami ◽  
S. Telmoune ◽  
...  
Keyword(s):  
Inert Matrix Fuel ◽  
Basic Properties ◽  
Inert Matrix ◽  
Iron Based

Behaviour of CS and I In Zirconia Based Fuel

1998 ◽  
Vol 540 ◽  
Author(s):  
C. Degueldre ◽  
M. Pouchon ◽  
M. Doebli ◽  
G. Ledergerber
Keyword(s):  
Incident Energy ◽  
Nuclear Reactors ◽  
Fission Products ◽  
Concentration Profiles ◽  
Reactor Safety ◽  
Inert Matrix Fuel ◽  
Light Water ◽  
Inert Matrix ◽  
Stepwise Heating ◽  
Implantation Depth

AbstractA zirconia based ceramic is foreseen as an inert matrix fuel for burning excess plutonium in light water nuclear reactors. For reactor safety reasons the behaviour of volatile fission products such as cesium and iodine must be studied since a retention of fission products is favourable for licensing the studied inert matrix fuel. In this study, implantation of Cs and I was performed into polycrystalline (Zr0.85, Y0.15)O1.925 samples. The implantation depth was selected on the basis of the ability to observe by Rutherford backscattering spectroscopy (RBS) the behaviour of Cs and I after treatment. With a 1 MeV incident energy, the ions are implanted at a depth of 200 nm as predicted by TRIM. After implantations full quantification of I and Cs concentration profiles was performed by RBS. The implantation profiles are measured as a function of sample temperature during stepwise heating programs. It is interesting to observe retention of Cs and I at relatively high temperature (e.g. for 2 h, below 900 K for Cs and below 1400 K for I). This behaviour is likely to be due to the size and interactions of these species in the zirconia solid solution.

Inert matrix fuel behaviour in test irradiations

2006 ◽  
Vol 352 (1-3) ◽  
pp. 291-299 ◽  
Author(s):  
Ch. Hellwig ◽  
M. Streit ◽  
P. Blair ◽  
T. Tverberg ◽  
F.C. Klaassen ◽  
...  
Keyword(s):  
Inert Matrix Fuel ◽  
Inert Matrix

Cermet sphere-pac concept for inert matrix fuel

2003 ◽  
Vol 319 ◽  
pp. 37-43 ◽  
Author(s):  
M.A. Pouchon ◽  
M. Nakamura ◽  
Ch. Hellwig ◽  
F. Ingold ◽  
C. Degueldre
Keyword(s):  
Inert Matrix Fuel ◽  
Inert Matrix

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).

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