Gap and Grain-Boundary Inventories of Cs, Tc, and Sr in Spent LWR Fuel

1991 ◽  
Vol 257 ◽  
Author(s):  
W. J. Gray ◽  
D. M. Strachan ◽  
C. N. Wilson
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Surface Area ◽  
Dissolution Behavior ◽  
Strongly Correlated ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Metallic Particles ◽  
Waste Container ◽  
Fission Gas Release

ABSTRACTSoluble radionuclides concentrated within the gap and grain-boundary regions of spent fuel could dissolve relatively rapidly were the waste container to fail and the fuel to be contacted by water in a geologic repository. To provide an estimate of the quantities of radionuclides that may be rapidly released, fractional inventories of Cs, Tc, and Sr concentrated within the fuel/cladding gap region have been measured for U.S. LWR spent fuels with fission gas release (FGR) values ranging from 0.25% to 18%. Separate measurements of the grain-boundary inventories of Cs, Tc, and Sr have been made for the same fuels. The Cs gap inventories were generally found to be about one fourth of the FGR values. The Cs grain-boundary inventories were generally less than 1% of the total Cs inventories and were not strongly correlated with FGR. Both the gap and grain-boundary inventories of Tc and Sr were near the detection limits of the methods used, less than 0.2% of the total inventories of these elements. However, some of the Tc may reside at the grain boundaries in the form of relatively insoluble metallic particles and not be detected by these experiments. Data obtained by comparing the dissolution behavior of fuel fragments with that of fuel grains were used to estimate the dissolution rate of Cs from the grain boundaries of one of the fuels. Surface-area normalized dissolution rates determined for fuel fragments in these same tests exceeded those determined for grains. A likely explanation is that the estimated fragment surface area did not take into account the “effective” grain-boundary surfaces.

Powder Leaching Study for Grain Boundary Inventory of Two High Burnup Fuels

MRS Advances ◽  
2019 ◽  
Vol 4 (17-18) ◽  
pp. 981-986
Author(s):  
Alexandre Barreiro Fidalgo ◽  
Olivia Roth ◽  
Anders Puranen ◽  
Lena Z. Evins ◽  
Kastriot Spahiu ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Spent Fuel ◽  
Short Period ◽  
Fission Gas ◽  
Leaching Solution ◽  
Significant Release ◽  
Leaching Experiments ◽  
High Burnup ◽  
Fission Gas Release

ABSTRACTIn the context of safety assessment, the fraction of inventory that is expected to rapidly dissolve when water contacts the spent fuel is called the Instant Release Fraction (IRF). Conceptually, this fraction consists of radionuclides outside of the uranium dioxide matrix and therefore the fraction can be further divided into the radionuclides in the fuel/cladding gap and radionuclides in the grain boundaries. The relative importance of these two fractions is investigated here for two Swedish high burnup fuels through simultaneous grinding and leaching fuel fragments in simplified groundwater for a short period of time. The hypothesis is that this will expose grain boundaries to leaching solution and provide an estimate of the release of the grain boundary inventory upon contact with water. The studied fragments were used in previous leaching experiments and thus pre-washed to remove any pre-oxidized phases. The results showed a significant release of iodine, cesium and rubidium and to a lower extent molybdenum and technetium. The fraction of inventory in the aqueous phase of actinides and lanthanides was 1-2 orders of magnitude lower than for the elements associated to the IRF. Both fuels displayed a very similar behavior and no correlation as a function of burnup or fission gas release was found.

Statistical Model for Grain Boundary and Grain Volume Oxidation Kinetics in UO2 Spent Fuel

1989 ◽  
Vol 176 ◽  
Author(s):  
R. B. Stout ◽  
H. F. Shaw ◽  
R. E. Einziger
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Density Function ◽  
Oxidation Kinetics ◽  
Yucca Mountain ◽  
Department Of Energy ◽  
Nuclear Waste Repository ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Integral Forms

ABSTRACTThe Yucca Mountain Project of the U.S. Department of Energy is investigating the suitability of a site in the unsaturated zone at Yucca Mountain, NV, for a high-level nuclear waste repository. Most of the waste will consist of UO2 spent fuel in Zircaloy-clad rods from nuclear reactors. If failure of both the waste containers and the cladding occurs within the lifetime of the repository, then the UO2 will be exposed to oxygen in the air and higher oxides of uranium may form. The oxidation state of the spent fuel may affect its dissolution behavior if later contacted by water. A model for the kinetics of spent fuel oxidation under repository-relevant conditions is thus necessary to predict the behavior of the waste form for assessing the performance of the repository with respect to the containment of radionuclides. In spent fuel experiments, the UO2 oxidation front initially propagates along grain boundaries followed by propagation into grain volumes. Thus, the oxidation kinetics is controlled by two processes and the oxidation of spent fuel fragments will depend on the density and physical attributes of grain boundaries. With this in mind, concepts from statistical mechanics are used to define a density function for grain boundaries perunit volume per unit species in a spent fuel fragment. Combining the integral forms of mass conservation and this grain boundary density function, a model for the global rate of oxidation for a spent fuel fragment is obtained. For rapid grain boundary oxidation compared to grain volume oxidation, equations of the model are solved and results compared to existing data.

The Importance of Criticality in the Safety Analysis of the Spent-Fuel Waste Container

1992 ◽  
Vol 294 ◽  
Author(s):  
William G. Culbreth ◽  
Paige Zielinski
Keyword(s):  
Thermal Effects ◽  
Federal Regulations ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Waste Container ◽  
Neutron Multiplication ◽  
Accident Conditions ◽  
High Level ◽  
Internal Configuration ◽  
Selection Of

ABSTRACTThe storage of high-level spent reactor fuel in a proposed national geologic repository will require the construction of containers to be placed in boreholes drilled into the host rock. Federal regulations require that the fuel be maintained subcritical under normal or accident conditions. This is determined through the calculation of a neutron multiplication factor, keff, that must remain below 0.95. Criticality will play an important role in the container design, the internal configuration of the fuel, and the selection of neutron poisons. An analysis of keff should be a normal step in the conceptualization of new waste container designs. Unlike thermal effects in a proposed repository, criticality will remain a problem long after the 10,000 year lifetime of the facility.

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.

Measurements of Grain-Boundary Inventories of 137Cs, 90Sr and 99Tc in Used Candu Fuel

1992 ◽  
Vol 294 ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
J.C. Tait ◽  
R.J. Porth ◽  
J.L. Mcconnell ◽  
T.R. Barnsdale ◽  
...  
Keyword(s):  
Aqueous Solution ◽  
Grain Boundary ◽  
Grain Boundaries ◽  
Source Term ◽  
Slow Release ◽  
Gas Release ◽  
System A ◽  
Fission Gas ◽  
Fission Gas Release ◽  
Upper Boundary

ABSTRACTTwo methods were used to measure grain-boundary inventories of 137Cs, 90Sr and 99Tc in used CANDU fuel, to corroborate source term estimates based on a fission gas release code. Used fuels were partially oxidized at 200°C in air to overall compositions of UO2+x (0.15≤ × ≤0.25) to expose UO2 grain boundaries, followed by leaching in aqueous solution. Only a fraction (2 to 18%) of the calculated gap + grain-boundary inventories for 37Cs was released. This suggests that the calculations overestimate Cs release or that oxidation does not expose all grain boundaries, or that Cs release from grain boundaries is slow. Release of 90Sr (0.01 to 0.7%) agreed reasonably well with the source term estimates (0.001 to 0.3%). Release of 99Tc (0.3 to 1.5%) suggests that the source term estimate for the upper boundary of 99Tc release (25%) may be too high. A second technique involved leaching of crushed and size-fractionated used fuel in either a static or dynamic system. A direct one-to-one correlation between calculated and measured gap + grain-boundary inventories for 137Cs was found for low- and medium-power fuels.

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.

scholarly journals Effects of Air Oxidation on the Dissolution Rate of LWR Spent Fuel

1992 ◽  
Vol 294 ◽  
Author(s):  
W. J. Gray ◽  
L. E. Thomas ◽  
R. E. Einziger
Keyword(s):  
Dissolution Rate ◽  
Surface Area ◽  
Solubility Limit ◽  
Oxidation States ◽  
Dissolution Mechanism ◽  
Coarse Particles ◽  
Spent Fuel ◽  
Air Oxidation ◽  
Geologic Repository ◽  
Practical Implications

ABSTRACTDissolution rates for air-oxidized spent fuel were measured in flowthrough tests where U concentrations were kept well below the solubility limit. Results from two types of specimens, separated grains and coarse particles, both in oxidized (U4O9+x) and unoxidized (UO2) conditions indicated only minor effects of oxidation on the surface-area-normalized rates. Similar results were obtained for unirradiated specimens in three different oxidation states (UO2, U3O7, and U3O8). These observations have important practical implications for disposal of spent fuel in a geologic repository as well as implications regarding the oxidative dissolution mechanism of UO2fuel.

Inventories of Iodine-129 and Cesium-137 in the Gaps and Grain Boundaries of LWR Spent Fuels

1999 ◽  
Vol 556 ◽  
Author(s):  
W. J. Gray
Keyword(s):  
Grain Boundary ◽  
Spent Nuclear Fuel ◽  
Yucca Mountain ◽  
Light Water Reactor ◽  
Gas Release ◽  
Spent Fuel ◽  
Fission Gas ◽  
Geologic Disposal ◽  
Cesium 137 ◽  
Fission Gas Release

AbstractPerformance assessment calculations that support geologic disposal of spent nuclear fuel in a potential repository at Yucca Mountain, Nevada, are based in part on the assumption that 2% of the total inventories of 135Cs, 129I, and 99Tc are located in the gap and grain-boundary regions where they could dissolve rapidly if the spent fuel were to be contacted by groundwater. Actual measured values reported here for a few light-water reactor (LWR) spent fuels show that the combined gap and grain-boundary inventories of 129I approximately equaled the fission-gas release fractions. For 137Cs, the combined gap and grain-boundary inventories were approximately one third of the fission-gas release fractions. These measured values can be used to replace the 2% estimate and thus reduce the uncertainties in the calculations.

Testing of Uranium Dioxide Containing Different Levels of Alpha Activity Under Anaerobic and Reducing Conditions

2008 ◽  
Vol 1104 ◽  
Author(s):  
Arvid Ödegaard-Jensen ◽  
Virginia Oversby
Keyword(s):  
Uranium Dioxide ◽  
Alpha Decay ◽  
Sample Surface ◽  
Hydrogen Atmosphere ◽  
Nitrogen Atmosphere ◽  
Alpha Activity ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Radiolysis Products

AbstractSweden plans to dispose of spent nuclear reactor fuel in a deep geologic repository in granitic rock. The conditions in the repository in the long term will be reducing and water is not expected to contact the fuel until after more than 1000 years. At that time, most of the beta- and gamma-active nuclides will have decayed away and the radiation will be dominated by alpha decay. In order to simulate the radiolysis field for dissolution of spent fuel with age more than 1000 years we have used uranium dioxide containing 5% U-235 and 0, 5, or 10% U-233. The 10% U-233 gives an alpha activity appropriate to about 3000 years after disposal. Samples were testied in a synthetic groundwater with low ionic strength and with the chemical composition dominated by sodium bicarbonate and calcium chloride. Tests were run in triplicate using an atmosphere of nitrogen (1atm), hydrogen (10 bar), hydrogen (10 bar) plus an iron strip in the solution, nitrogen (1 atm) plus an iron strip in the solution, hydrogen (10 bar) plus an iron strip in the solution, hydrogen (10 bar) without the iron strip. Each of these test conditions was run for 2 consecutive periods of at least 21 days. The results showed that the dissolution behavior of the samples was the same for both nitrogen atmosphere and hydrogen atmosphere. The amount of U dissolved under these conditions clearly showed the enhancement of dissolution due to oxidation of the sample surface by radiolysis products. When an iron strip was added to the solution, the amount of dissolution decreased dramatically indicating that the Fe(II) ions released from the corroding iron were able to react with most of the radiolysis products before they could oxidize the uranium dioxide surface.

Measurement of the Displacement Across a Coincidence Grain Boundary

1977 ◽  
Vol 35 ◽  
pp. 124-125
Author(s):  
J. W. Matthews ◽  
W. M. Stobbs
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Stacking Fault ◽  
The Other ◽  
Measuring Technique ◽  
High Angle ◽  
Twin Boundaries ◽  
Rigid Displacement ◽  
Cubic Crystals ◽  
Lattice Displacement

Many high-angle grain boundaries in cubic crystals are thought to be either coincidence boundaries (1) or coincidence boundaries to which grain boundary dislocations have been added (1,2). Calculations of the arrangement of atoms inside coincidence boundaries suggest that the coincidence lattice will usually not be continuous across a coincidence boundary (3). There will usually be a rigid displacement of the lattice on one side of the boundary relative to that on the other. This displacement gives rise to a stacking fault in the coincidence lattice.Recently, Pond (4) and Smith (5) have measured the lattice displacement at coincidence boundaries in aluminum. We have developed (6) an alternative to the measuring technique used by them, and have used it to find two of the three components of the displacement at {112} lateral twin boundaries in gold. This paper describes our method and presents a brief account of the results we have obtained.

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