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.

Coprecipitation Studies from Simfuel Solutions in 5 m NaCl Media

1996 ◽  
Vol 465 ◽  
Author(s):  
P. Diaz-Arocas ◽  
J. Garcia-Serrano
Keyword(s):  
Fission Products ◽  
Solution Concentration ◽  
Laser Fluorescence ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Minor Elements ◽  
Icp Ms ◽  
Upper Limits ◽  
Ph Range ◽  
Irradiated Fuel

ABSTRACTExtensive Research is performed in many countries in order to evaluate the spent fuel behaviour under repository conditions. Several aspects as the control of the oxidative spent fuel dissolution by secondary phases formation are not yet clear.Coprecipitation experiments from SIMFUEL solutions are performed to study if minor elements will influence the formation of secondary phases. Therefore, coprecipitation studies from SIMFUEL solutions aims at identification of stable phases of significant simulated fission products. These experiments provide upper limits for solution concentration and distribution ratios of simulate fission products at several pH values. SIMFUEL pellets, which simulate an irradiated fuel with burnup of 50 GWd/tU were provided by AECL Research Laboratories, Canada. Experiments were carried out by addition of an aliquot of the initial SIMFUEL solution to 5 m NaCI free of carbonates solution. The selected pH was maintained constant during the experiments. The pH range considered was from 5.5 to 9.3. Analyses of the solutions were performed for uranium by Laser fluorescence and for the minor elements by ICP-MS. Solid phases formed at pH 5.5 were dissolved and analysed by ICP-MS. Results of the evolution in solution vs. pH of simulated fission products concentrations are shown in this paper.

Xps and Sem Studies on the Corrosion of UO2 Cointaining Plutonium in Demineralized and Carbonated Water.

2002 ◽  
Vol 757 ◽  
Author(s):  
J. Cobos ◽  
T. Wiss ◽  
T. Gouder ◽  
V. V. Rondinella
Keyword(s):  
Radiation Field ◽  
Surface Oxidation ◽  
Photoemission Spectroscopy ◽  
Spent Fuel ◽  
Experimental Conditions ◽  
Different Types ◽  
Leaching Experiments ◽  
Carbonated Water ◽  
Storage Times

ABSTRACTAn oxidation and dissolution study has been performed on UO2 pellets containing ∼10 and ∼0.1 wt. % 238Pu, ∼10 wt. % 239Pu and on undoped UO2 to investigate the effects of radiolysis and composition on the corrosion behavior of spent fuel. The so-called alpha-doped UO2 is used to simulate the alpha-radiation field of different types of commercial LWR spent fuel after different storage times. Leaching experiments in demineralized and carbonated water at room temperature under oxidizing conditions showed that relatively high amounts of 238Pu were released. The leached surfaces were examined with X-ray Photoemission Spectroscopy (XPS), and the progressive surface oxidation was monitored. The oxidation of the U(IV) during the leaching experiments, in the materials doped with 238Pu resulted in precipitation of U(VI) phases: enhanced formation of studtite for the strongest radiation field and shoepite at low radiation field was observed on the surface of the pellet. Essentially no precipitation of Pu-rich phases was directly observed. Leaching in carbonated water and characterization of UO2 containing 239Pu under the same experimental conditions were performed and the results compared to those for alpha-doped UO2. The chemistry effects due to the presence of Pu in addition to alpha-radiolysis were investigated.

Simfuel Dissolution Studies in Granitic Groundwater at 25°C

1990 ◽  
Vol 212 ◽  
Author(s):  
Amaia Sandino ◽  
I. Casas ◽  
K. Ollila ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Minor Component ◽  
Minor Components ◽  
Spent Fuel ◽  
Cumulative Release ◽  
Dissolution Studies ◽  
Leaching Experiments ◽  
Granitic Groundwater ◽  
Chemical Analogue

ABSTRACTThe dissolution of a non-radioactive chemical analogue of spent nuclear fuel (SIMFUEL) has been studied as a function of two different synthetic groundwaters at 25°C. The cumulative release of U, Mo, Ba, Y and Sr is presented after 170 days of total leaching time.The results obtained show the potential usefulness of SIMFUEL to ascertain the kinetics and mechanisms of dissolution of the minor components of spent fuel. In the case of Sr, a good correlation is found with the dissolution of this minor component measured in spent fuel leaching experiments.

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

2008 ◽  
Vol 1107 ◽  
Author(s):  
F. Clarens ◽  
D. Serrano-Purroy ◽  
A. Martínez-Esparza ◽  
D. Wegen ◽  
E. Gonzalez-Robles ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Correction Factor ◽  
Core Sample ◽  
Gas Release ◽  
Spent Fuel ◽  
Fission Gas ◽  
Leaching Experiments ◽  
Fission Gas Release ◽  
Fractional Release ◽  
Number Of Particles

AbstractThe so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.

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.

Effect of Secondary Phases Formation in the Leaching of UO2 Under Simulated Radiolytic Products

1994 ◽  
Vol 353 ◽  
Author(s):  
P. Díaz-Arocas ◽  
J. Quinoñes ◽  
C. Maffiotte ◽  
J. Serrano ◽  
J. Garcia ◽  
...  
Keyword(s):  
Secondary Phase ◽  
Point Of View ◽  
Spent Fuel ◽  
Secondary Phases ◽  
X Ray Diffraction ◽  
X Ray ◽  
Ph Values ◽  
Leaching Process ◽  
Leaching Experiments ◽  
Solid Phases

AbstractThe leaching of the spent fuel matrix (UO2) is function of the radiolytic products formation. The effect of each radioiytic product on the leaching process is not totally understood. In the literature, the influence of H2O2 on the dissolution process is described from the qualitative point of view, and most of the studies were performed for pH values from 8 to 12. In this paper we report on the effect of the H2O2 in the leaching process of UO2 by dissolution experiments at various H2O2 concentrations. Also, it was tested the influence of S/V ratio (surface area exposed to the leaching media) on the UO2 leaching and secondary phases formation. It was identified the formation of secondary phases on the UO2 surface. Solid phases characterization was carried out by x-ray Photoelectron Spectrometry (XPS), x-ray Diffraction (XRD) and Scanning Electron Microscopy (SEM) techniques. By XPS studies the secondary phase formed corresponded to a U(VI) phase. By XRD analyses the solid was identified as studtite, UO4 - 4H2O, (card n0 16–206, [I]). A comparison of the U(VI) phases formed in spent fuel and UO, leaching experiments in various media has been carried out.

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.

Sign in / Sign up

Export Citation Format

Share Document