scholarly journals Interactive leach tests of UO/sub 2/ and spent fuel with waste package components in salt brine

1986 ◽  
Author(s):  
J.O. Barner ◽  
W.J. Gray ◽  
G.L. McVay ◽  
J.W. Shade
Keyword(s):  
Spent Fuel ◽  
Waste Package ◽  
Salt Brine

Migration Rates of Brine Inclusions in Single Crystals of Nacl

1981 ◽  
Vol 6 ◽  
Author(s):  
I-Ming Chou
Keyword(s):  
Radioactive Waste ◽  
Temperature Gradients ◽  
Spent Fuel ◽  
Ion Diffusion ◽  
Cold Side ◽  
Migration Rates ◽  
Permanent Storage ◽  
Salt Brine ◽  
Salt Deposits ◽  
High Level

Rock-salt deposits have been considered as a possible medium for the permanent storage of high-level radioactive wastes and spent fuel. Brine inclusions present in natural salt can migrate toward the waste if the temperature and the temperature gradients in the vicinity of the radioactive waste are large enough. This migration is due to the dissolution of salt at the hot side of the salt-brine interface, ion diffusion through the brine droplet, and the precipitation of salt at the cold side of the salt brine interface.

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.

The Expected Environment for Waste Packages in a Salt Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
L. R. Pederson ◽  
D. E. Clark ◽  
F. N. Hodges ◽  
G. L. Mcvpy ◽  
D. Rai
Keyword(s):  
Rock Salt ◽  
Near Field ◽  
Materials Testing ◽  
Site Specific ◽  
Waste Package ◽  
Barrier Materials ◽  
Waste Forms ◽  
Temperature Irradiation ◽  
Salt Brine ◽  
Package Materials

ABSTRACTThis paper discusses results of recent efforts to define the very near-field (within approximately 2m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and “standard” brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported.

Evaluation of Spent Fuel as a Waste Form In a Salt Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
Walter J. GRAY ◽  
Gary L. McVay ◽  
John. O. Barner ◽  
John W. Shade ◽  
Roger W. Cote
Keyword(s):  
Deionized Water ◽  
Waste Form ◽  
Spent Fuel ◽  
Permian Basin ◽  
Salt Brine

ABSTRACTLeach tests have been performed on spent fuel in synthetic Permian Basin salt brine at 25 and 75°C. Complementary tests on unirradiated UO2 pellets have been conducted in both salt brine and deionized water in the range 25 to 150°C. Iron and/or oxidized zircaloy coupons were included in some of the tests. Uranium release from spent fuel was more than 100 times greater than from U02. In brine, iron had no significant effect on the total uranium release but substantially reduced the amount in solution by causing the uranium to plate out on the iron coupon and container walls and to precipitate as filterable particles.

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.

Analysis Of Congruent Matrix Release, Precipitation, And Time-Distributed Containment Failure On Spent Fuel Performance

1987 ◽  
Vol 112 ◽  
Author(s):  
Michael J. Apted ◽  
David W. Engel
Keyword(s):  
Source Term ◽  
Spent Fuel ◽  
Release Rates ◽  
Waste Package ◽  
Regulatory Limits ◽  
Lower Solubility ◽  
Fractional Abundance ◽  
Congruent Dissolution ◽  
Cesium 137 ◽  
Term Evaluation

AbstractThe Analytical Repository Source-Term (AREST) code has been developed for source-term evaluation of spent fuel as a final waste form in geologic repositories. AREST contains a set of analytical equations for the timedependent diffusional mass transport of both solubility-limited and inventory-limited radionuclides from a spent fuel in a failed container surrounded by a shell of packing or other porous material imbedded in a porous host rock. Three factors that affect release performance are examined: 1) congruent dissolution of the UO2 matrix, 2) chemical instability of the UO2 matrix, with precipitation of a more stable uranium phase within the waste package, and 3) the attenuation of release rate by distribution of containment failures with time.For congruent matrix dissolution, the release rates of included nuclides are proportional to the product of solubility-limited release of uranium and the fractional abundance of the nuclide. For certain conditions, congruent release rates are calculated to be up to 10 orders of magnitude lower than release rates assuming individual solubility-limits. Precipitation of a more stable, lower solubility uranium phase within the waste package is shown to increase release rates from the UO2 matrix compared to the non-precipitation case, in agreement with previous calculations. During the first 300 to 1000 years after repository closure, the distribution of containment failures with time will act to attenuate the peak average release rates of soluble, longlived nuclides, such as iodine-129, to values smaller than release rates below regulatory limits. However, for soluble nuclides with short half-lives, such as cesium-137, a broader distribution of containment failure with constant mean time of failure can actually cause an increase In the peak average release rates.

scholarly journals Consequence modelling of hypothetical post-closure criticality events for spent fuel disposal

2015 ◽  
Vol 79 (6) ◽  
pp. 1505-1513 ◽  
Author(s):  
R. M. Mason ◽  
J. K. Martin ◽  
P. N. Smith ◽  
R. J. Winsley
Keyword(s):  
International Studies ◽  
Quasi Steady State ◽  
Spent Fuel ◽  
Regulatory Requirement ◽  
Geological Disposal ◽  
Waste Package ◽  
Decay Heat ◽  
Disposal Facility ◽  
Safety Case ◽  
The Uk

AbstractIn support of the Radioactive Waste Management (RWM) safety case for a geological disposal facility (GDF) in the UK, there is a regulatory requirement to consider the likelihood and consequences of nuclear criticality. Waste packages are designed to ensure that criticality is not possible during the transport and operational phases of a GDF and for a significant period post-closure. However, over longer post-closure timescales, conditions in the GDF will evolve.For waste packages containing spent fuel, it can be shown that, under certain conditions, package flooding could result in a type of criticality event referred to as 'quasi-steady-state' (QSS). Although unlikely, this defines a 'what-if' scenario for understanding the potential consequences of post-closure criticality. This paper provides an overview of a methodology to understand QSS criticality and its application to a spent fuel waste package.The power of such a hypothetical criticality event is typically estimated to be a few kilowatts: comparable with international studies of similar systems and the decay heat for which waste packages are designed. This work has built confidence in the methodology and supports RWM's demonstration that post-closure criticality is not a significant concern.

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