scholarly journals Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

1998 ◽  
Author(s):  
W.J. Gray ◽  
R.E. Einziger
Keyword(s):  
Dissolution Rate ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Rate Testing ◽  
Aqueous Conditions ◽  
Initial Results

scholarly journals Comparison and Modeling of Aqueous Dissolution Rates of Various Uranium Oxides

1996 ◽  
Vol 465 ◽  
Author(s):  
S. A. Steward ◽  
E. T. Mones
Keyword(s):  
Dissolution Rate ◽  
Atmospheric Oxygen ◽  
Fold Increase ◽  
Stable Phase ◽  
Uranium Oxides ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Rates ◽  
Intrinsic Dissolution ◽  
Carbonate Concentration

ABSTRACTThe purpose of this work has been to measure and model the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a geologic repository. When exposed to air at elevated temperature, spent fuel may form the stable phase U3O8. Dehydrated schoepite, UO3H2O, has been shown to exist in drip tests on spent fuel.Equivalent sets of U3O8 and UO3H2 dissolution experiments allowed a systematic examination of the effects of temperature (25–75°C), pH (8–10) and carbonate (2–200×10−4 molar) concentrations at atmospheric oxygen conditions.Results indicate that UO3H2O has a much higher dissolution rate (at least ten-fold) than U3O8 under the same conditions. The intrinsic dissolution rate of unirradiated U3O8 is about twice that of UO2. Dissolution of both U3O8 and UO3.H2O shows a very high sensitivity to carbonate concentration. Present results show a 25 to 50-fold increase in room-temperature UO3H2O dissolution rates between the highest and lowest carbonate concentrations.As with the UO2 dissolution data the classical observed chemical kinetic rate law was used to model the U3O8 dissolution rate data. The pH did not have much effect on the models, in agreement with the earlier analysis of the UO2 and spent fuel dissolution data,. However, carbonate concentration, not temperature, had the strongest effect on the U3O8 dissolution rate. The U3O8 dissolution activation energy was about 6000 cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and UO2 respectively.

Initial Results From Dissolution Testing of Spent Fuel Under Acidic Conditions

2004 ◽  
Vol 824 ◽  
Author(s):  
Brady D. Hanson ◽  
Judah I. Friese ◽  
Chuck Z. Soderquist
Keyword(s):  
Reaction Rates ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Tests ◽  
Acidic Conditions ◽  
Ph Conditions ◽  
Long Term Behavior ◽  
Initial Results ◽  
Instantaneous Release

AbstractFlowthrough dissolution tests using solutions with pH in the range 2 to 7 have been conducted on a moderate burnup Light Water Reactor spent fuel. Such low pH conditions have been modeled as possibly occurring in a failed waste package at the proposed repository at Yucca Mountain. The release oftotal uranium, 99Tc, 90Sr, 137Cs, and 239&240Pu were measured for up to 90% total reaction of the specimens. The reaction rates, determined both from the cumulative release and the release normalized to surface area, were found to decrease with increasing pH and with increasing extent of reaction. The implications to instantaneous release and long-term behavior ina geologic repository are discussed.

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.

Phase Relations of the Uranyl Oxide Hydrates and their Relevance to the Disposal of Spent Fuel

1990 ◽  
Vol 212 ◽  
Author(s):  
R. J. Finch ◽  
R. C. Ewing
Keyword(s):  
General Formula ◽  
Phase Relations ◽  
Molecular Water ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Structural Sheet ◽  
Natural Analogues ◽  
Ph Conditions ◽  
Image Position

ABSTRACTUranyl oxide hydrates, formed by the alteration of uraninite, are natural analogues for the long-term corrosion products of spent fuel in a geologic repository under oxidizing conditions. The uranyl oxide hydrates may be represented by the general formula:Pb-bearing hydrates require the addition of a neutral uranyl group into the structural sheet (UO2(OH)2) for each interlayer Pb ion. Distortion of the structure associated with the additional uranyl groups is reduced by replacing two structural hydroxyls with a structural oxygen and a molecular water. The general formula for the Pb-uranyl oxide hydrates is:This hypothesis explains the paragenetic sequences:1) schoepite ➛ billietite ➛ protasite ➛ bauranoite2) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ masuyite ➛ wölsendorfite3) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ ± masuyite ➛ sayrite ➛ curite, and indicates that, under relatively high pH conditions, schoepite will not be the long-term solubility-controlling phase for uranium in uranium-rich groundwaters.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

scholarly journals The Role Of Chemical Reaction In Waste-Form Performance

1987 ◽  
Vol 112 ◽  
Author(s):  
P. L. Chambré ◽  
C. H. Kang ◽  
W. W.-L. Lee ◽  
T. H. Pigford
Keyword(s):  
Mass Transfer ◽  
Steady State ◽  
Dissolution Rate ◽  
Chemical Reaction ◽  
Borosilicate Glass ◽  
Reaction Rate ◽  
Waste Form ◽  
Spent Fuel ◽  
Wide Range ◽  
Chemical Reaction Rate

AbstractThe dissolution rate of waste solids in a geologic repository is a complex function of waste form geometry, chemical reaction rate, exterior flow field, and chemical environment. We present here an analysis to determine the steady-state mass transfer rate, over the entire range of flow conditions relevant to geologic disposal of nuclear waste. The equations for steady-state mass transfer with a chemical-reaction-rate boundary condition are solved by three different mathematical techniques which supplement each other. This theory is illustrated with laboratory leach data for borosilicate-glass and a spherical spent-fuel waste form under typical repository conditions. For borosilicate glass waste in the temperature range of 57°C to 250°C, dissolution rate in a repository is determined for a wide range of chemical reaction rates and for Peclet numbers from zero to well over 100, far beyond any Peclet values expected in a repository. Spent-fuel dissolution in a repository is also investigated, based on the limited leach data now available.

A Comparison of the Performance of Nuclear Waste Glasses by Modeling

1987 ◽  
Vol 112 ◽  
Author(s):  
B. Grambow ◽  
D. M. Strachan
Keyword(s):  
Nuclear Waste ◽  
Nuclear Reactors ◽  
Liquid Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Geologic Repository ◽  
Waste Forms ◽  
The Earth ◽  
Soluble Solid ◽  
High Level

The reprocessing of spent fuel from nuclear reactors and processing of fuels for defense purposes have generated large volumes of high-level liquid waste that need to be immobilized prior to final storage. For immobilization, the wastes must be converted to a less soluble solid, and, although other waste forms exist, glass currently appears to be the choice for the transuranic-containing portion of the reprocessed waste. Once produced, this glass will be sent in canisters to a geologic repository located some 200 to 500 m below the surface of the earth.

The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository

1993 ◽  
Vol 333 ◽  
Author(s):  
William G. Culbreth ◽  
Paige R. Zielinski
Keyword(s):  
Fuel Burnup ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Chain Reaction ◽  
Worst Case ◽  
Mixture Ratio ◽  
Waste Package ◽  
Radiological Data ◽  
A Chain ◽  
High Level

ABSTRACTStudies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.

Sign in / Sign up

Export Citation Format

Share Document