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.

Mechanisms for the Formation of a Perched Water Zone in Fractured Tuff: a Natural Analogue Study

1996 ◽  
Vol 465 ◽  
Author(s):  
E. G. Woodhouse ◽  
R. L. Bassett
Keyword(s):  
Structural Features ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Arid Environments ◽  
Perched Water ◽  
Natural Analog ◽  
Welded Tuff ◽  
Geochemical Properties ◽  
Water Zone ◽  
Mountain Site

ABSTRACTPerched water zones have been identified in the fractured, welded tuff in the semi-arid to arid environments of Yucca Mountain, Nevada and near Superior, Arizona. An understanding of the formation of such zones is necessary in order to predict where future perched water might form at Yucca Mountain, the proposed site of a high-level nuclear waste repository. The formation or growth of a perched zone near a repository is one of the factors to be considered in the risk assessment of the Yucca Mountain site.The Apache Leap Research Site near Superior, Arizona is a natural analog to the Yucca Mountain site in terms of geology, hydrology, and climate. Information used to study possible mechanisms for the formation of the perched zone included data regarding isotopie and geochemical properties of the waters in and above the perched water zone; measured hydrologie parameters of the perched zone; geophysical and measured parameters of the tuff; megascopic and microscopic observations of the tuff, including mineralogical, alteration, and structural features; and the lateral and vertical extent of perched water in the region.Aquifer test, geophysical, geochemical, and radioisotopic data show that fractures are the means by which water is recharging the perched zone. The reduced hydraulic conductivity of the formation in the perched zone appears to result from both a severe reduction in matrix porosity and permeability caused by welding, devitrification, and vapor phase crystallization; and by an increase in fracture filling which restricts the pathways for flow.

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.

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.

Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on Radionuclide Release

2010 ◽  
Author(s):  
Lubna K. Hamdan ◽  
John C. Walton ◽  
Arturo Woocay
Keyword(s):  
Heat Release ◽  
Nuclear Waste ◽  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Minimum Length ◽  
General Corrosion ◽  
Waste Package ◽  
Flow Through ◽  
Radionuclide Release ◽  
Residual Heat

Over time, nuclear waste packages disposed in geological repositories are expected to fail gradually due to localized and general corrosion. As a result, water will have access to the nuclear waste and radionuclides will be transported to the accessible environment by ground water. In this paper we consider a serious failure case in which penetrations at the top and bottom of the waste package will allow water to flow through it (flow-through model). We introduce a new conceptual model that examines the effect of the residual heat release of the nuclear waste stored in an unsaturated environment on radionuclide release. This model predicts that the evaporation of water at the hotter sheltered areas (from condensate and seepage) inside the failed waste package will create a capillary pressure gradient that drives water to wick with its dissolved and suspended contents toward these relict areas, effectively preventing radionuclides release. We drive a dimensionless group to estimate the minimum length of the sheltered areas required to sequester radionuclides and prevent their release. The implications of this model on the performance of the proposed repository at Yucca Mountain or unsaturated zone geological repositories in general are explored.

Alternate Source Term Models for Yucca Mountain Performance Assessment Based on Natural Analog Data and Secondary Mineral Solubility

1999 ◽  
Vol 556 ◽  
Author(s):  
William M. Murphy ◽  
Richard B. Codell
Keyword(s):  
Performance Assessment ◽  
Release Rate ◽  
Source Term ◽  
Yucca Mountain ◽  
Solution Chemistry ◽  
Base Case ◽  
Total System ◽  
Spent Fuel ◽  
Natural Analog ◽  
Mineral Solubility

AbstractPerformance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Peña Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Peña Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg yr−1, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Peña Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.

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.

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

scholarly journals Ion Exchange and Dehydration Effects on Potassium and Argon Contents of Clinoptilolite

1995 ◽  
Vol 412 ◽  
Author(s):  
G. Woldegabriel ◽  
S. Levy
Keyword(s):  
Unsaturated Zone ◽  
Yucca Mountain ◽  
Nuclear Waste Repository ◽  
Radionuclide Transport ◽  
Geochemical Processes ◽  
Saturated Zone ◽  
High Level Nuclear Waste ◽  
Waste Repository ◽  
Natural Clinoptilolite ◽  
High Level

AbstractZeolite-rich Miocene tuffs are an important part of the principal hydrochemical barrier to water-borne radionuclide transport from a potential high-level nuclear waste repository at Yucca Mountain, Nevada. The timing of zeolitization is an issue that relates to paleohydrology, permeability, zeolite stability, and unsaturated-zone geochemical processes. Exploratory K/Ar dating of clinoptilolite, the most abundant and widespread zeolite, shows a striking and consistent pattern of increasing apparent ages (2–13 Ma) with depth. Only the isotopic ages from the saturated zone are compatible with geologic evidence suggesting an age >10 Ma for most of the zeolites.Factors that may be responsible for the young apparent ages in the unsaturated zone were investigated. Cation exchange with recharge water and Ar diffusion under unsaturated conditions (processes that may be characteristic of the unsaturated zone) were evaluated experimentally for their effects on K/Ar systematics. Cation exchanging a natural clinoptilolite with Ca-, Cs-, K-, and Na-chloride solutions showed minimal effects on radiogenic Ar content. However, clinoptilolite heated at 2007deg;C for 16 hours in air lost a significant amount of its radiogenic Ar compared with minimal losses from clinoptilolite heated in water at 100°C for over 5 months. The preliminary results indicate that Ar loss from incompletely hydrated clinoptilolite may be a major factor contributing to the young apparent ages of clinoptilolite in the unsaturated zone at Yucca Mountain.

The Dissolution of Uranophane in CaCl2-SiO2(aq) Test Solutions

2004 ◽  
Vol 824 ◽  
Author(s):  
James D. Prikryl ◽  
William M. Murphy
Keyword(s):  
Yucca Mountain ◽  
Ore Deposits ◽  
Nuclear Waste Repository ◽  
Chemical Analyses ◽  
Spent Fuel ◽  
Dissolution Reaction ◽  
Batch Dissolution ◽  
Siliceous Rocks ◽  
Waste Repository

AbstractUranophane [Ca(UO2)2(SiO3OH)2 · 5H2O] is a corrosion product of long-term leaching of spent fuel under oxidizing conditions and is a weathering product of uraninite in uranium ore deposits hosted by siliceous rocks. Incorporation of radionuclides into uranophane by coprecipitation may occur as a result of spent fuel alteration. Dissolution of uranophane leading to release of these radionuclides may therefore influence the longterm dissolved concentration and mobility of radionuclides at the proposed nuclear waste repository at Yucca Mountain, Nevada. In this study, the dissolution of uranophane in Ca- and Si-rich test solutions was investigated. Batch dissolution experiments were designed to approach uranophane equilibrium from undersaturated solutions at nearneutral pH (~6.0). Test solutions had initial U concentrations of 0.0 and 10-7 mol/L in matrices of ~10-2 mol/L CaCl2 and ~10-3 mol/L SiO2(aq). The test solutions were reacted with synthetic uranophane (confirmed by XRD and chemical analyses) and analyzed periodically over 10 weeks. Reaction quotients (Log Qs) derived from aqueous speciations of experimental solutions considered to be near equilibrium with uranophane ranged from 10.54 to 11.06 for the dissolution reaction: Ca(UO2)2(SiO3OH)2 · 5H2O + 6H+ ⇔ Ca2+ + 2UO22+ + 2SiO2(aq) + 9H2O.

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