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.

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.

scholarly journals An Investigation into Dissolution Rate of Spent Nuclear Fuel in Aqueous Reprocessing

2004 ◽  
Vol 41 (2) ◽  
pp. 126-134 ◽  
Author(s):  
Hideaki MINEO ◽  
Hikaru ISOGAI ◽  
Yasuji MORITA ◽  
Gunzo UCHIYAMA
Keyword(s):  
Dissolution Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel

Fast/Instant Radionuclide Release: Effects inherent to the experiment

2014 ◽  
Vol 1665 ◽  
pp. 233-243
Author(s):  
B. Kienzler ◽  
A. Loida ◽  
E. González-Robles ◽  
N. Müller ◽  
V. Metz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Solution Chemistry ◽  
Release Rates ◽  
Radionuclide Sorption ◽  
Series Of Experiments ◽  
Radionuclide Release

ABSTRACTThe release of radionuclides measured during washing cycles of spent nuclear fuel samples in a series of experiments using different solutions are analyzed with respect to the fission products Cs, Sr, and Tc and the actinides U, Pu, and Am. Based on the concentrations of the dissolved radionuclides, their release rates are evaluated in terms of fraction of inventory in the aquatic phase per day. The application of this information on the fast/instant radionuclide release fraction (IRF) is discussed and following issues are addressed: Duration of the wash steps, solution chemistry, and radionuclide sorption onto surface of the experimental vessels. Data for the IRF are given and the correlation between the mobilization of the various elements is analyzed.

scholarly journals Author Correction: Distribution of metallic fission-product particles in the cladding liner of spent nuclear fuel

2020 ◽  
Vol 4 (1) ◽  
Author(s):  
Richard A. Clark ◽  
Michele A. Conroy ◽  
Timothy G. Lach ◽  
Edgar C. Buck ◽  
Kristi L. Pellegrini ◽  
...  
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel

scholarly journals Distribution of metallic fission-product particles in the cladding liner of spent nuclear fuel

2020 ◽  
Vol 4 (1) ◽  
Author(s):  
Richard A. Clark ◽  
Michele A. Conroy ◽  
Timothy G. Lach ◽  
Edgar C. Buck ◽  
Kristi L. Pellegrini ◽  
...  
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel

Characterization of selenium in UO2spent nuclear fuel by micro X-ray absorption spectroscopy and its thermodynamic stability

2015 ◽  
Vol 17 (10) ◽  
pp. 1760-1768 ◽  
Author(s):  
E. Curti ◽  
A. Puranen ◽  
D. Grolimund ◽  
D. Jädernas ◽  
D. Sheptyakov ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Fission Product ◽  
Nuclear Fuel ◽  
Absorption Spectroscopy ◽  
Thermodynamic Stability ◽  
Spent Nuclear Fuel ◽  
X Ray ◽  
Geological Repository ◽  
X Ray Absorption

The long-lived fission product79Se is tightly bound to the UO2lattice in spent nuclear fuel; it will thus be released only very slowly from a geological repository for radioactive waste.

79Se: Geochemical and Crystallo-Chemical Retardation Mechanisms

1999 ◽  
Vol 556 ◽  
Author(s):  
Fanrong Chen ◽  
Peter C. Burns ◽  
Rodney C. Ewing
Keyword(s):  
Fission Product ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Structural Connectivity ◽  
Structural Similarity ◽  
High Solubility ◽  
Alkaline Conditions ◽  
Ph Conditions ◽  
Selenium Analysis ◽  
Aqueous Species

Abstract79Se is a long-lived (1.1×106 years) fission product which is chemically and radiologically toxic. Under Eh-pH conditions typical of oxidative alteration of spent nuclear fuel, selenite or selenate are the dominant aqueous species of selenium. Because of the high solubility of metalselenites and metal-selenates and the low adsorption of selenite and selenate aqueous species under alkaline conditions, selenium may be highly mobile. However, 79Se released from altered fuel may be immobilized by incorporation into secondary uranyl phases as low concentration impurities, and this may significantly reduce the mobility of selenium. Analysis and comparison of the known structures of uranyl phases indicate that (SeO3) may substitute for (SiO3OH) in structures with the uranophane anion-topology (α.-uranophane, sklodowskite, boltwoodite) which are expected to be the dominant alteration phases of UO2 in Si-rich groundwater. The structural similarity of guillemninite, Ba[(UO2)3 (SeO3)2O2](H2O) 3 to phurcalite, [(UO2)3(PO4)2O2](H2O)7, suggests that the substitution (SeO3)↔ (PO4) may occur in phurcalite. The close similarity between the sheets in the structures of rutherfordine and [(UO2)(SeO3)] implies that the substitution (SeO3) ↔ (CO3) can occur in rutherfordine. However, the substitutions: (SeO3) ↔ (SiO3OH) in soddyite and (SeO3) ↔ (PO4) in phosphuranylite may disrupt their structural connectivity and are unlikely to occur.

The effects of alpha-radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu-doped UO2

2002 ◽  
Vol 90 (9-11) ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
F. King ◽  
J. S. Betteridge ◽  
F. Garisto
Keyword(s):  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Rates ◽  
Term Stability ◽  
Long Term Stability ◽  
Alpha Radiolysis

SummaryThe long-term stability of spent nuclear fuel under deep geologic repository conditions will be determined mostly by the influence of α-radiolysis, since the dose-rate for α-radiolysis will exceed that for γ/β-radiolysis beyond a fuel age of ∼100 years and will persist for more than 10000 years. Dissolution rates derived from studies with currently available spent fuel include radiolysis effects from γ/β- as well as α-radiolysis. The use of external α-sources and chemically added H

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