Determination of Criteria for Selecting A UO2Fuel Dissolution Model for Nuclear Fuel Waste Management Concept Assessment

1991 ◽  
Vol 257 ◽  
Author(s):  
S. Sunder ◽  
D.W. Shoesmith ◽  
N.H. Miller ◽  
G.J. Wallace
Keyword(s):  
Kinetic Model ◽  
Dissolution Rate ◽  
Nuclear Fuel ◽  
Electrochemical Techniques ◽  
Experimental Studies ◽  
Solution Chemistry ◽  
Redox Conditions ◽  
Water Radiolysis ◽  
Oxidative Dissolution ◽  
Dissolution Rates

ABSTRACTAssessing the concept of direct disposal of used nuclear fuel in a geological vault requires a model to predict the dissolution rate of UO2in groundwater. A solubility-limited model can be used to calculate the dissolution rate of UO2fuel under non-oxidizing conditions. When the oxidative dissolution of UO2is an irreversible process, a kinetic model is more suitable to describe the dissolution of UO2under oxidizing conditions. Experimental studies were carried out using electrochemical techniques and X-ray photoelectron spectroscopy, XPS, to determine criteria for selecting the appropriate model for estimating used-fuel dissolution rates as a function of the redox conditions in the vault at the time of container failure. UO2electrodes were subjected to prolonged (>1000 min) potentiostatic oxidation, and the rate of oxidation and dissolution of UO2fuel was investigated as a function of the applied potential. UO2oxidation was also carried out by the products of water radiolysis and studied as a function of dose rate, total dose and solution chemistry.These studies show that significant oxidative dissolution of UO2appears possible for potentials more positive than -100 mV vs SCE in solutions with a pH close to that of the deep groundwaters, i.e., from 6 to 10. A kinetic model, which takes into account the mechanism of UO2oxidation, is more appropriate to estimate dissolution rates of UO2fuel for redox conditions more oxidizing than -100 mV vs SCE.

Corrosion Study of SIMFUEL in Aerated Carbonate Solution Containing Calcium and Silicate

2013 ◽  
Vol 1518 ◽  
pp. 139-144
Author(s):  
Hundal Jung ◽  
Tae Ahn ◽  
Roberto Pabalan ◽  
David Pickett
Keyword(s):  
Fission Product ◽  
Dissolution Rate ◽  
Nuclear Fuel ◽  
Polarization Resistance ◽  
Spent Nuclear Fuel ◽  
Electrochemical Impedance ◽  
Major Effect ◽  
Solution Chemistry ◽  
High Rate ◽  
Dissolution Rates

ABSTRACTThe corrosion behavior of simulated spent nuclear fuel (SIMFUEL) was investigated using electrochemical impedance spectroscopy and solution chemistry analyses. The SIMFUEL was exposed to aerated solutions of NaCl+NaHCO3 with and without calcium (Ca) and silicate. Two SIMFUEL compositions were studied, representing spent nuclear fuel (SNF) corresponding to 3 or 6 at % burnup in terms of fission product equivalents of surrogate elements. For all tested cases, the polarization resistance increased with increased immersion time, indicating possible blocking effects due to accumulation of corrosion products on the SIMFUEL surface. The potential-pH diagram suggests formation of schoepite that may cause the increase in the polarization resistance. The addition of Ca and silicate produced no measureable change in the polarization resistance measured at the corrosion potential. The dissolution rate ranged from 1 to 3 mg/m2-day, which is similar to the range of dissolution rates for SIMFUEL and SNF reported in the literature for comparable conditions. SIMFUEL burnup did not have a major effect on the dissolution rate. Analysis of the solution chemistry shows that uranium is the dominant element dissolved in the posttest solutions, and the dissolution rates calculated from uranium (U) concentrations are consistent with the dissolution rates obtained from impedance measurements. Simulated-fission product elements (i.e., barium, molybdenum, strontium, and zirconium) dissolved from the SIMFUEL electrode at a relatively high rate. Sorption test results indicated significant sorption of U onto the oxide formed on stainless steel. Electrochemical methods were found to be effective for measuring the uranium dissolution rate in real time.

Uranium (iv) Dioxide and Simfuel as Chemical Analogues of Nuclear Spent Fuel Matrix Dissolution. A Comparison of Dissolution Results in a Standard Naci/NaHCO3 Solution

1994 ◽  
Vol 353 ◽  
Author(s):  
Jordi Bruno ◽  
I. Casas ◽  
E. Cera ◽  
J. de Pablo ◽  
J. GimÉnez ◽  
...  
Keyword(s):  
Dissolution Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Dissolution Process ◽  
Experimental Comparison ◽  
Comparison Study ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Surface Sites ◽  
Radiolytic Oxidation

AbstractWe have carried out an experimental comparison study of the dissolution rates of unirradiated UO2 and SIMFUEL pellets and particles (100–300 μm) in a standard NaCI/NaHC03 solution, under oxidizing conditions. We have performed the experiments using batch and flow methodologies. Both methodologies gave similar results, indicating that the overall oxidation/dissolution process is the same in both cases. The results from the experiments indicate that under these conditions the dissolution process is both oxygen and bicarbonate promoted. The dissolution rates we obtained are: R=2.4 ± 0.8 mg U/m2 d for U02 and R= 0.17 ± 0.05 mg U/m2 d for SIMFUEL. The results of the experiments indicate that the dissolution rate under oxic conditions is clearly dependent on the number of U(VI) surface sites which for spent nuclear fuel is a function of the extent of radiolytic oxidation.

scholarly journals A Kinetic Model for Borosilicate Glass Dissolution Based on the Dissolution Affinity of a Surface Alteration Layer

1989 ◽  
Vol 176 ◽  
Author(s):  
William L. Bourcier ◽  
Dennis W. Peiffer ◽  
Kevin G. Knauss ◽  
Kevin D. McKeegan ◽  
David K. Smith
Keyword(s):  
Kinetic Model ◽  
Dissolution Rate ◽  
Borosilicate Glass ◽  
Geochemical Modeling ◽  
Secondary Phases ◽  
Dissolution Rates ◽  
Ion Concentrations ◽  
Glass Dissolution ◽  
Gel Layer ◽  
Amorphous Surface

ABSTRACTA kinetic model for the dissolution of borosilicate glass, incorporated into the EQ3/6 geochemical modeling code, is used to predict the dissolution rate of a nuclear waste glass. In the model, the glass dissolution rate is controlled by the rate of dissolution of an alkalidepleted amorphous surface (gel) layer. Assuming that the gel layer dissolution affinity controls glass dissolution rates is similar to the silica saturation concept of Grambow [1] except that our model predicts that all components concentrated in the surface layer, not just silica, affect glass dissolution rates. The good agreement between predicted and observed elemental dissolution rates suggests that the dissolution rate of the gel layer limits the overall rate of glass dissolution. The model predicts that the long-term rate of glass dissolution will depend mainly on ion concentrations in solution, and therefore on the secondary phases which precipitate and control ion concentrations.

Prediction of the Oxidative Dissolution Rates of Used Nuclear Fuel in a Geological Disposal Vault Due to the Alpha Radiolysis of Water

1994 ◽  
Vol 353 ◽  
Author(s):  
S. Sunder ◽  
D.W. Shoesmith ◽  
N.H. Miller
Keyword(s):  
Nuclear Fuel ◽  
Management Program ◽  
Oxidative Dissolution ◽  
Dissolution Rates ◽  
Used Nuclear Fuel ◽  
A Value ◽  
Thin Layer Cell ◽  
Alpha Radiolysis ◽  
Nuclear Fuel Waste ◽  
Electrochemical Model

AbstractEffects of alpha radiolysis of water on the corrosion of nuclear fuel (UO2) have been investigated in solutions at pH = 9.5, i.e., a value close to that expected in groundwaters at the depth of the disposal vault proposed in the Canadian nuclear fuel waste management program, CNFWMP. The corrosion potentials of UO2 electrodes exposed to the products of alpha radiolysis of water were monitored as a function of alpha flux and exposure time in a specially designed thin-layer cell. The oxidative dissolution rates of UO2 are calculated from the steady-state values of the corrosion potential using an electrochemical model. A procedure to predict the dissolution rate of used nuclear fuel in groundwater as a function of fuel cooling time is described, and illustrated by calculating the dissolution rates for the reference used fuel in the CNFWMP (Bruce CANDU reactor fuel, burnup 685 GJ/kg U). It is shown that the oxidative dissolution of used fuel in the CNFWMP will be important only for time periods ≤ 600 a at this burnup and assuming no decrease in pH.

From Laboratory Experiments to a Geological Disposal Vault: Calculation of Used Nuclear Fuel Dissolution Rates

1997 ◽  
Vol 506 ◽  
Author(s):  
S. Sunder ◽  
D.W. Shoesmith ◽  
M. Kolar ◽  
D.M. Leneveu
Keyword(s):  
Waste Management ◽  
Nuclear Fuel ◽  
Management Program ◽  
Gamma Dose ◽  
Redox Conditions ◽  
Dissolution Rates ◽  
Geological Disposal ◽  
Dose Rates ◽  
Used Nuclear Fuel ◽  
Gamma Dose Rates

ABSTRACTCalculation of used nuclear fuel dissolution rates in a geological disposal vault requires a knowledge of the redox conditions in the vault. For redox conditions less oxidizing than those causing UO2 oxidation to the U3O7 stage, a thermodynamically-based model is appropriate. For more oxidizing redox conditions a kinetic or an electrochemical model is needed to calculate these rates. The redox conditions in a disposal vault will be affected by the radiolysis of groundwater by the ionizing radiation associated with the fuel. Therefore, we have calculated the alpha-, beta- and gamma-dose rates in water in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) as a function of cooling time. Also, we have determined dissolution rates of UO2 fuel as a function of alpha and gamma dose rates from our electrochemical measurements. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100°C, the highest temperature expected in a container in the CNFWMP, as a function of time since emplacement. It is shown that beta radiolysis of water will be the main cause of oxidation of used CANDU fuel in a failed container. The use of a kinetic or an electrochemical corrosion model, to calculate fuel dissolution rates, is required for a period of ∼ 1000 a following emplacement of copper containers in the geologic disposal vault envisaged in the CNFWMP. Beyond this time period a thermodynamically-based model adequately predicts the fuel dissolution rates. The results presented in this paper can be adopted to calculate used fuel dissolution rates for other used UO2 fuels in other waste management programs.

The reductive dissolution of synthetic goethite and hematite in dithionite

Clay Minerals ◽  
1987 ◽  
Vol 22 (3) ◽  
pp. 329-337 ◽  
Author(s):  
J. Torrent ◽  
U. Schwertmann ◽  
V. Barron
Keyword(s):  
Specific Surface ◽  
Dissolution Rate ◽  
Surface Area ◽  
Reductive Dissolution ◽  
Natural Environments ◽  
Dissolution Rates ◽  
Aluminium Substitution ◽  
Initial Dissolution Rate

AbstractThe reductive dissolution by Na-dithionite of 28 synthetic goethites and 26 hematites having widely different crystal morphologies, specific surfaces and aluminium substitution levels has been investigated. For both minerals the initial dissolution rate per unit of surface area decreased with aluminium substitution. At similar aluminium substitution and specific surface, goethites and hematites showed similar dissolution rates. These results suggest that preferential, reductive dissolution of hematite in some natural environments, such as soils or sediments, might be due to the generally lower aluminium substitution of this mineral compared to goethite.

scholarly journals A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

2021 ◽  
Vol 13 (3) ◽  
pp. 1442
Author(s):  
Sanggil Park ◽  
Jaeyoung Lee ◽  
Min Bum Park
Keyword(s):  
Kinetic Model ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Complete Loss ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Loss Of Coolant ◽  
Model Derivation ◽  
Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

scholarly journals Properties of the high burnup structure in nuclear light water reactor fuel

2017 ◽  
Vol 105 (11) ◽  
Author(s):  
Thierry Wiss ◽  
Vincenzo V. Rondinella ◽  
Rudy J. M. Konings ◽  
Dragos Staicu ◽  
Dimitrios Papaioannou ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Fuel ◽  
Experimental Studies ◽  
Fuel Pellet ◽  
Extended Defects ◽  
Safety Margins ◽  
Fission Gas ◽  
High Burnup ◽  
Formed Structure

AbstractThe formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings from extensive studies performed at JRC-Karlsruhe during the last 25 years to determine properties and behaviour of the HBS. In particular, microstructural features, thermal transport, fission gas behaviour, and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental studies is that the HBS does not compromise the safety of nuclear fuel during normal operations.

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