Coffinitization - A Mechanism for the Alteration of UO2 under Reducing Conditions

1991 ◽  
Vol 257 ◽  
Author(s):  
J. Janeczek ◽  
R.C. Ewing
Keyword(s):  
Fission Products ◽  
Spent Fuel ◽  
Dissolved Silica ◽  
Reducing Conditions ◽  
Chemical Analogue

ABSTRACTA structural and chemical analogue of UO2 in spent fuel, uraninite, UO2+x, is unstable in the presence of dissolved silica (>10-3.6 mg/L) under reducing conditions and transforms into coffinite, USiO4 nH20. Coffinite may incorporate numerous elements into its structure including actinides and fission products. Conditions favorable for coffinitization of UO2 in spent fuel may occur in repositories in granitic, basaltic, and tuffaceous rocks.

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.

Leaching of Spent Fuel under Anaerobic and Reducing Conditions

2002 ◽  
Vol 757 ◽  
Author(s):  
Yngve Albinsson ◽  
Arvid Ödegaard-Jensen ◽  
Virginia M. Oversby ◽  
Lars O. Werme
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Fluid Phase ◽  
Inert Atmosphere ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Reducing Conditions ◽  
Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

Leaching Study of the Behaviour of Spent Fuel and SIMFUEL Under Simulated Granitic Repository Conditions

2001 ◽  
Author(s):  
J. A. Serrano ◽  
J. Quiñones ◽  
J. Cobos ◽  
P. Diaz Arocas ◽  
V. V. Rondinella ◽  
...  
Keyword(s):  
Fission Products ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Experimental Conditions ◽  
Radiation Fields ◽  
Leaching Experiments ◽  
Irradiated Fuel ◽  
Fractional Release ◽  
Intense Radiation ◽  
Chemical Analogue

Abstract The leaching behaviour of spent fuel is of importance for the concept of direct storage of spent fuel. The aim of this study was to study UO2 irradiated fuel under simulated granitic repository conditions. In parallel with these spent fuel tests, SIMFUEL leaching studies were also performed. Direct comparisons between spent fuel and its chemical analogues, SIMFUEL, are often difficult. On one hand, because of the differences existing between spent fuel and SIMFUEL. E.g., for irradiated fuel: different origin and burnup, presence of intense radiation fields, hence radiolysis effects, or formation of cracks and pores due to the volatile fission products, hence larger surface area. On the other hand, because of different experimental procedures used by different authors. This work presents results of sequential leaching experiments in synthetic granite water in equilibrium with a cylinder of granite at room temperature in air using spent UO2 fuel and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. The specimens used were UO2 (43 MWd/kgU) and SIMFUEL (simulating a burnup of 30 MWd/kgU) as non-irradiated chemical analogue. A thermodynamic study by means of geochemistry codes was also performed. Differences both in fractional release and in uranium concentration in the leachate were found. The highest fractional release of uranium was measured for UO2 spent fuel. Candidate solid phases calculated for controlling the uranium solubility were soddyite ((UO2)2(SiO4)·2H2O) in the case of spent fuel and haiweete (Ca(UO2)2(Si2O5)3·5H2O) for SIMFUEL. Further work is ongoing to characterise the surfaces of the leached fuel samples and to try to confirm the preliminary attempts to identify reprecipitated secondary phases. Comparison of some fission product release between spent fuel and SIMFUEL was also performed.

The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell
Keyword(s):  
Low Carbon Steel ◽  
Fission Products ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Waste Package ◽  
Dissolved Species ◽  
Radionuclide Release ◽  
Time Invariant

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.

Spent Fuel: Characterization Studies and Dissolution Behaviourunder Disposal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
L. H. Johnson ◽  
D. W. Shoesmith ◽  
S. Stroes-Gascoyne
Keyword(s):  
Release Kinetics ◽  
Fission Products ◽  
Solution Chemistry ◽  
Future Research ◽  
Spent Fuel ◽  
Research Areas ◽  
Dissolution Model ◽  
Underground Disposal ◽  
Radionuclide Release

AbstractThe concept of disposal of unreprocessed spent fuel has now been under study internationally for over ten years. Considerable progress has been made in understanding the factors that will control radionuclide release from spent fuel in an underground disposal vault. This progress is reviewed and the research areas of significance in providing further data for source term models are discussed. Key areas for future research are identified; these include improved characterization of spent fuel to determine the inventories of fission products at grain boundaries, together with their release kinetics; and a better understanding of the effects of solution chemistry on spent fuel dissolution, in particular the effects of salinity, redox chemistry, and radiolysis of groundwater. Approaches to modelling the dissolution of spent fuel are discussed, and a possible approach for developing an oxidative dissolution model is outlined.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

Nondestructive, Energy-Dispersive, X-Ray Fluorescence an Analysis of Actinide Stream Concentrations from Reprocessed Nuclear Fuel

1979 ◽  
Vol 23 ◽  
pp. 163-176
Author(s):  
D. C. Camp ◽  
W. D. Ruhter
Keyword(s):  
Nuclear Fuel ◽  
Nitrate Solution ◽  
Fission Products ◽  
Analytical Techniques ◽  
Fluorescence Analysis ◽  
Energy Dispersive ◽  
Light Water Reactors ◽  
Spent Fuel ◽  
X Ray ◽  
Water Reactors

In the event that nuclear fuel from light water reactors (LWR) is reprocessed to reclaim the uranium or plutonium, several analytical techniques will be used for product accountability. Generally, the isotopic content of both the plutonium and uranium in the reprocessed product will have to be accurately determined. One plan for the reprocessing of LWR spent fuel incorporates the following scheme. After separation from both the fission products and transplutonium actinides (including neptunium and americium), part of the uranium and all of the plutonium in a nitrate solution will merge together to form a coprocessed stream. This solution will be concentrated by evaporation and sent to a hold tank for accountability. Input concentrations into the hold tank could be up to 350 g U/ℓ and nearly 50 g Pu/ℓ. The variation to be expected in these concentrations is not known. The remaining uranium fraction will be further purified and sent to a separate storage tank. Its expected stream concentration will be about 60 g U/ℓ. These two relatively high actinide stream concentrations can be monitored rapidly, quantitatively, and nondestructively using the technique of energy-dispersive x-ray fluorescence analysis(XRFA).

Operation Experience of HALW Storage Process at the Tokai Reprocessing Plant

2001 ◽  
Author(s):  
Yoshinobu Nakamura ◽  
Shizuka Suda ◽  
Koich Ishiyama ◽  
Masaru Watahiki ◽  
Hideyo Mutoh
Keyword(s):  
Liquid Waste ◽  
Fission Products ◽  
Storage Management ◽  
Storage Tanks ◽  
Spent Fuel ◽  
Storage Process ◽  
Highly Active ◽  
Radioactive Nuclides ◽  
Active Liquid ◽  
Reprocessing Plant

Abstract Volume of nitric acid solution, which contain the most of fission products (FPs), is concentrated to 0.5–2m3 in an HALW evaporator by reprocessing a spent fuel of ltU. The HALW is stored in HALW storage tanks temporarily till it is transferred to the Tokai Vitrification Facility (TVF). During the storage, the HALW of 1–8m3/y vaporized per storage tank. A shift coefficient of radioactive nuclides from the HALW to off-gas was 4E−11. Through the operation experience, knowledge obtained about storage management of highly active liquid waste (HALW) is reported.

Corrosion performance of ferrous and refractory metals in molten salts under reducing conditions

1999 ◽  
Vol 14 (5) ◽  
pp. 1990-1995 ◽  
Author(s):  
J. E. Indacochea ◽  
J. L. Smith ◽  
K. R. Litko ◽  
E. J. Karell
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Lithium Oxide ◽  
Liquid Lithium ◽  
Chloride Salt ◽  
Spent Fuel ◽  
Ferrous Alloys ◽  
Materials Development ◽  
Reducing Conditions ◽  
Lithium Reduction

A lithium reduction technique to condition spent fuel for disposal has been developed at the Argonne National Laboratory. There is a need to ensure adequate vessel longevity through corrosion testing and, if necessary, materials development. Several ferrous alloys and tantalum specimens were submitted to a corrosion test at 725 °C for thirty days in an argon atmosphere, using a lithium-chloride salt saturated with lithium metal and containing small amounts of lithium oxide and lithium nitride. The samples did not show dimensional or weight change, nor could corrosion attack be detected metallographically. The lithium-saturated salt system did not show any behavior similar to that of liquid lithium corrosion. From testing in other gas compositions, it appears that the presence of oxygen in the system is necessary to produce severe corrosion.

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