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.

scholarly journals Effects of Gamma Radiolysis on Waste Package Components

1986 ◽  
Vol 84 ◽  
Author(s):  
Michele A. Lewis ◽  
Donald T. Reed
Keyword(s):  
Carbon Steel ◽  
Gamma Radiation ◽  
Low Carbon Steel ◽  
Pressure Vessels ◽  
Ion Concentration ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Waste Package ◽  
Gamma Radiolysis ◽  
Synthetic Groundwater

AbstractPreliminary short-term experiments to study the effect of gamma radiation on the waste package materials interactions in the environment of a nuclear waste repository in basalt were completed. Experimental parameters were similar to those expected during the saturated post- closure period of the repository, i.e., a temperature of 200°C and a hydrostatic pressure of ≤70 bar. The test components included various combinations of synthetic groundwater (GR4), low carbon steel coupons, CH4, basalt, and basalt/bentonite mixtures (packing). The tests were run in low carbon steel pressure vessels.It was found that gamma radiation at a dose rate of 104rad/hr for a duration of one and two months increased the H2 yield in all tests. Increases in the organic carbon yield, the sulfate ion concentration, and the corrosion rate of the coupon were observed in some of the tests. These latter results varied with the combination of waste package components included in the tests. Evidence for the quench effect was observed in tests which included basalt.

Spent Fuel: Characterization Studies and Dissolution Behaviourunder Disposal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
L. H. Johnson ◽  
D. W. Shoesmith ◽  
S. Stroes-Gascoyne
Keyword(s):  
Release Kinetics ◽  
Fission Products ◽  
Solution Chemistry ◽  
Future Research ◽  
Spent Fuel ◽  
Research Areas ◽  
Dissolution Model ◽  
Underground Disposal ◽  
Radionuclide Release

AbstractThe concept of disposal of unreprocessed spent fuel has now been under study internationally for over ten years. Considerable progress has been made in understanding the factors that will control radionuclide release from spent fuel in an underground disposal vault. This progress is reviewed and the research areas of significance in providing further data for source term models are discussed. Key areas for future research are identified; these include improved characterization of spent fuel to determine the inventories of fission products at grain boundaries, together with their release kinetics; and a better understanding of the effects of solution chemistry on spent fuel dissolution, in particular the effects of salinity, redox chemistry, and radiolysis of groundwater. Approaches to modelling the dissolution of spent fuel are discussed, and a possible approach for developing an oxidative dissolution model is outlined.

Reactions in the System Basalt/Simulated Spent Fuel/Water

1983 ◽  
Vol 26 ◽  
Author(s):  
D.E. Grandstaff ◽  
G.L. Mckeon ◽  
E.L. Moore ◽  
G.C. Ulmer
Keyword(s):  
Water System ◽  
Potassium Feldspar ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Glass Dissolution ◽  
Ph Values ◽  
C 30 ◽  
High Level ◽  
Radionuclide Release

ABSTRACTThe Grande Ronde Basalts underlying the Hanford Site are being evaluated as a possible site for a high-level nuclear waste repository. Experiments, in which basalt from the Umtanun flow of the Grande Ronde Basalt and basalt with simulated spent fuel were reacted with synthetic Hanford groundwater, were conducted to determine steady state concentrations which can be used in radionuclide release-rate models. Tests were performed at temperatures of 100°, 200°, and 300°C; 30 MPa pressure, and a solution:solid mass ratio of 10:1 for durations up to 7,000 hr. Solution aliquots were extracted periodically during the experiments for analysis. The pH was measured at 250°C and recalculated to higher temperatures. In the basalt-water system the stable high-temperature pH values achieved were 7.2 (100°C), 7.5 (200°C), and 7.6 (300°C). Solution composition variations are due to mesostasis (glass) dissolution and precipitation of secondary phases. Solution measurements indicate a redox potential (Eh) of about -0.7 volts at 300°C. Secondary phases produced include silica, potassium feldspar, iron oxides, clays, scapolite, and zeolites. Tests in the basalt + simulated spent fuel + water systen show that calculated pH values stabilized near 7.6 (100°C), 7.2 (200°C), and 7.7 (300°C). At higher temperatures, solution concentrations were controlled by secondary phases similar to those found in basalt-water tests. Less than 1% of uranium, thorium, samarium, rhenium, cerium, and palladium were released to solution while somewhat higher amounts of iodine, molybdenum, and cesium were released. The UO2 component was unreactive; however, other components (e.g., cesium-bearing phases) were almost completely dissolved. Secondary phases incorporating radionuclide-analog elements include clays, palladium sulfide, powellite, coffinite, and a potassium-uranium silicate.

Characterization of LWR Spent Fuel Approved Testing Materials for Radionuclide Release Studies

1984 ◽  
Vol 44 ◽  
Author(s):  
John O. Barner ◽  
J. L. Daniel
Keyword(s):  
Fission Products ◽  
Materials Characterization ◽  
Light Water Reactor ◽  
Spent Fuel ◽  
Light Water ◽  
Water Reactor ◽  
Release Studies ◽  
Radionuclide Release

AbstractThe Materials Characterization Center (MCC) provides characterized light-water reactor (LWR) spent fuel approved testing materials (ATMs) for use by experimenters in radionuclide release studies in support of repository licensing efforts. The characterization completed to date is described for MCC ATM-101, a Zircaloy-clad UO2 fuel from the H. B. Robinson, Unit 2, Assembly BO-5. ATM-101 is a moderate burnup, low-releasing (fission products in-reactor) spent fuel typical of the fuel that is expected to initially be placed in a repository.

Hot Cell Preparation of Tesiing Materials

1984 ◽  
Vol 44 ◽  
Author(s):  
R. E. Thornhill ◽  
C. A. Knox
Keyword(s):  
Fission Products ◽  
Materials Characterization ◽  
High Level Waste ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Cell Preparation ◽  
High Radiation ◽  
High Level ◽  
Commercial Reactor ◽  
Simulated High Level Waste

AbstractIt is important in nuclear waste repository development that testing be done with materials containing a radionuclide spectrum representative of actual wastes. To meet the need for such materials, the Materials Characterization Center (MCC) has prepared simulated high level waste (HLW) glasses with radionuclides representative of about 10-, 300-, and 1000-year-old waste. A quantity of well characterized spent fuel also has been acquired for the same purpose. Glasses containing 10- and 300-year-old wastes, and the spent fuel specimens, must be fabricated in a hot cell. Hot cell conditions (high radiation field, remote operation, and difficulty of repairs) require that procedures and equipment normally used in materials preparation out-of-cell be modified for hot cell applications.This paper discusses the fabrication of two glasses, and the preparation of test specimens of these glasses and spent fuel. One of the glasses is a 76–68 composition, which is fully loaded with actual commercial reactor fission product waste. The other glass contains simulated Barnwell Nuclear Fuel Plant waste, doped with different combinations of fission products and actinides. The spent fuel is a 10-year-old PWR material. Special techniques have been used to achieve high quality, well characterized testing materials, including specimens in the form of segments, wafers, cylinders, and powders of these materials.

Unsaturated Zone Waters From the Nopal I Natural Analog, Chihuahua, Mexico - Implications for Radionuclide Mobility at Yucca Mountain

1999 ◽  
Vol 556 ◽  
Author(s):  
David A. Pickett ◽  
William M. Murphy
Keyword(s):  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Path Modeling ◽  
Rock Interaction ◽  
Colloidal Transport ◽  
Natural Analog ◽  
Initial Silica ◽  
Radionuclide Release

AbstractChemical and U-Th isotopic data on unsaturated zone waters from the Nopal I natural analog reveal effects of water-rock interaction and help constrain models of radionuclide release and transport at the site and, by analogy, at the proposed nuclear waste repository at Yucca Mountain. Geochemical reaction-path modeling indicates that, under oxidizing conditions, dissolution of uraninite (spent fuel analog) by these waters will lead to eventual schoepite precipitation regardless of initial silica concentration provided that groundwater is not continuously replenished. Thus, less soluble uranyl silicates may not dominate the initial alteration assemblage and keep dissolved U concentrations low. Uranium-series activity ratios are consistent with models of U transport at the site and display varying degrees of leaching versus recoil mobilization. Thorium concentrations may reflect the importance of colloidal transport of low-solubility radionuclides in the unsaturated zone.

scholarly journals Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage

2019 ◽  
Vol 5 (1) ◽  
pp. 61-66
Author(s):  
Artem Z. Gayazov ◽  
Sergey V. Komarov ◽  
Anton Yu. Leshchenko ◽  
Ksenia E. Revenko ◽  
Valery P. Smirnov ◽  
...  
Keyword(s):  
Fission Product ◽  
Hydrogen Generation ◽  
Accumulation Rate ◽  
Fission Products ◽  
Spent Fuel ◽  
Abrupt Decrease ◽  
Gaseous Fission ◽  
Gaseous Fission Product ◽  
Radionuclide Release ◽  
Product Accumulation

The paper describes the outcomes of the experiments to study hydrogen and gaseous fission products accumulation during simulations of the wet damaged VVER-440 SNF storage in air-tight canisters with the water drained and no drying conducted. Physical and chemical processes occurring during the damaged oxide SNF storage in wet environment are discussed. The experiments were carried out in two stages: 1) preliminary soaking of fine fuel particles in water in an air-tight canister, 2) water draining and keeping the wet SNF in the air-tight canister. The experiments were conducted one after another using the same SNF canister and differing only in the SNF soaking temperature, i.e. 25 and 80 °С. The radionuclide release into the liquid phase during the SNF storage under water was studied. Uranium and cesium isotopic concentrations were found to reach steady values when the SNF is kept under water for more than a month. The kinetics of hydrogen and gaseous fission product accumulation in the gaseous phase during wet storage of the spent fuel in the air-tight canister with the water drained coincide for both experiments. The kinetics demonstrate an abrupt decrease of the hydrogen and gaseous fission product accumulation rate in 46 hours. The data obtained can be applied for development and verification of the damaged SNF behavior models during SNF storage in wet environment under radiolysis.

Effect of Spent Fuel Burnup and Composition on Alteration of the U(Pu)O2 Matrix

2000 ◽  
Vol 663 ◽  
Author(s):  
C. Jégou ◽  
S. Peuget ◽  
J.F. Lucchini ◽  
C. Corbel ◽  
V. Broudic ◽  
...  
Keyword(s):  
Release Kinetics ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Static Mode ◽  
Mox Fuel ◽  
The Matrix ◽  
Leaching Experiments ◽  
Experimental Findings ◽  
Radionuclide Release ◽  
Kinetics Of

ABSTRACTFor a potential performance assessment of direct disposal of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified.Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd·tHM−1 and a MOX fuel sample irradiated to 47 GWd·tHM−1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25°C) in a hot cell.After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 à 10−7 d−1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings.

Incorporation of Radioactive Metals and Trace Metals into Phosphate “Anhydrous” Autunite: Thermodynamic Evaluation

2002 ◽  
Vol 757 ◽  
Author(s):  
Huifang Xu ◽  
Yifeng Wang
Keyword(s):  
Environmental Remediation ◽  
Order Approximation ◽  
Fission Products ◽  
Monovalent Cation ◽  
Nuclear Waste Repository ◽  
Linear Free Energy Relationship ◽  
Thermodynamic Evaluation ◽  
Spent Fuel ◽  
Free Energies ◽  
Linear Free Energy

ABSTRACTThe prediction of heavy metal solubility in the presence of phosphate is the key for designing a successful environmental remediation strategy or for the performance assessment of a nuclear waste repository. In order to evaluate the possibility for solubilities and incorporation of U and its fission products, such as Cs, Np, Pu, it is important to have the Gibbs free energies of formation of possible end-member phases. In this paper, we use a linear free energy relationship to calculate the Gibbs free energies of formation of divalent cation “anhydrous” autunite phases [M2+(UO2)2(PO4)2] and monovalent cation “anhydrous” autunite phases [M+2(UO2)2(PO4)2]. Based on predicted Gibbs free energies of formation for autunite phases, an equilibrium calculation indicates that it is feasible to incorporate NpO2+ into the autunite phases. If concentration of Cs+ is high, Cs+ will be incorporated into the autunite phases. However, it is not feasible to incorporate low concentration Cs+ released from spent fuel into the uranyl phosphate phases with layered structure and interlayer water (such as autunite phases, boltwoodite phases, and uranophane phases) as previously proposed. The predicted data can be considered as a first order approximation.

Rim Instant Release Radionuclide Inventory From French High Burnup Spent UOX Fuel

2009 ◽  
Vol 1193 ◽  
Author(s):  
D. Roudil ◽  
C. Jegou ◽  
V. Broudic ◽  
M. Tribet
Keyword(s):  
Small Volume ◽  
Source Term ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Waste Container ◽  
High Burnup ◽  
Radionuclide Release ◽  
Long Term Behavior ◽  
The Impact

AbstractTo assess the long-term behavior of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. The instant release fraction (IRF) source term assumed to be instantaneously accessible to water after the failure of the waste container. Some IRF values for different kinds of spent fuel are available in the literature. However, the possible contribution the rim restructured zone for high-burnup UOX fuels, was not necessary taken into account. A specific study of the leaching behavior of the rim zone has been carried out on a UOX fuel sample with a burnup of 60 GWd·t-1 and 2.8% FGR. The 134/137Cs and 90Sr rim IRF are effectively higher than the gap and grain boundary inventories (respectively 4.4 wt% and 0.3 wt% of the total Cs and Sr inventories). Nevertheless because of the relative small volume of this zone in the pellet, the impact of the rim inventory appears to be limited and the complete Cs and Sr IRF (gap, grain boundaries and rim), were estimated at 1.85 and 0.3 wt%, respectively

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