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.

UO2 dissolution in high pH conditions of the Belgian Supercontainer

2012 ◽  
Vol 1475 ◽  
Author(s):  
Th. Mennecart ◽  
C. Cachoir ◽  
K. Lemmens
Keyword(s):  
Spent Fuel ◽  
Dissolution Rates ◽  
High Ph ◽  
Neutral Ph ◽  
Dissolution Tests ◽  
Alkaline Conditions ◽  
Ph Conditions ◽  
Ph Range ◽  
Long Term Behavior

ABSTRACTTo assess the long-term behavior of spent fuel in alkaline conditions representative for the Belgian Supercontainer design, static and dynamic dissolution tests were performed with depleted and Pu-doped UO2 , simulating medium burn-up UOX fuels of different fuel ages. The experiments were performed under argon atmosphere at 25 – 30°C in cement waters in the pH range 11.7 – 13.5 and at different SA/V ratios. This paper presents the observed UO2 matrix dissolution rates based on the (238U or 233U) release, and proposes a selection of reference dissolution rates for performance assessment. We demonstrate that the dissolution rates at high pH are equivalent to the dissolution rates reported in the literature for neutral pH conditions. The α-activity threshold below which radiolytical fuel oxidation becomes negligible, seems to be close to the threshold reported for anoxic media at neutral pH.

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.

Dissolution of spent nuclear fuel fragments at high alkaline conditions under H2 overpressure

MRS Advances ◽  
2016 ◽  
Vol 1 (62) ◽  
pp. 4163-4168
Author(s):  
E. González-Robles ◽  
M. Herm ◽  
V. Montoya ◽  
N. Müller ◽  
B. Kienzler ◽  
...  
Keyword(s):  
Aqueous Solution ◽  
Partial Pressure ◽  
Nuclear Fuel ◽  
Alkaline Solution ◽  
Spent Nuclear Fuel ◽  
Dissolution Tests ◽  
The Matrix ◽  
Alkaline Conditions ◽  
Long Term Behavior

ABSTRACTThe long-term behavior of the UO2 fuel matrix under conditions of the Belgian “Supercontainer design” was investigated by dissolution tests of high burn-up spent nuclear fuel (SNF) in high alkaline solution under 40 bar of (Ar + 8%H2) atmosphere. Four fragments of SNF, obtained from a pellet previously leached during two years, were exposed to young cement water with Ca (YCWCa) under 3.2 bar H2 partial pressure in four single/independent autoclave experiments for a period of 59, 182, 252 and 341 days, respectively. After a decrease of the concentration of dissolved 238U, which is associated with a reduction of U(VI) to U(IV), the concentration of 238U in solution is constant in the experiments running for 252 and 341 days. These observations indicate an inhibition of the matrix dissolution due to the presence of H2. A slight increase in the concentration of 90Sr and 137Cs in the aqueous solution indicates that there is still dissolution of the grain boundaries. These findings are similar to those reported for spent nuclear fuel corrosion in synthetic near neutral pH solutions.

Static dissolution tests of α-doped UO2 under repository relevant conditions: Influence of Boom Clay and α-activity on fuel dissolution rates

2006 ◽  
Vol 932 ◽  
Author(s):  
Sonia Salah ◽  
Christelle Cachoir ◽  
Karel Lemmens ◽  
Norbert Maes
Keyword(s):  
Clay Fraction ◽  
Dissolution Kinetics ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Retention Capacity ◽  
Boom Clay ◽  
Dissolution Tests ◽  
Clay Formation ◽  
Α Activity

ABSTRACTSince reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ~263 µg · m−2 · d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.

The effects of alpha-radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu-doped UO2

2002 ◽  
Vol 90 (9-11) ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
F. King ◽  
J. S. Betteridge ◽  
F. Garisto
Keyword(s):  
Dose Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Rates ◽  
Term Stability ◽  
Long Term Stability ◽  
Alpha Radiolysis

SummaryThe long-term stability of spent nuclear fuel under deep geologic repository conditions will be determined mostly by the influence of α-radiolysis, since the dose-rate for α-radiolysis will exceed that for γ/β-radiolysis beyond a fuel age of ∼100 years and will persist for more than 10000 years. Dissolution rates derived from studies with currently available spent fuel include radiolysis effects from γ/β- as well as α-radiolysis. The use of external α-sources and chemically added H

scholarly journals Corroded spent nuclear fuel examined with EELS

1996 ◽  
Vol 54 ◽  
pp. 562-563
Author(s):  
Edgar C. Buck ◽  
Nancy L. Dietz ◽  
John K. Bates
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Yucca Mountain ◽  
Radiological Hazard ◽  
Geologic Repository ◽  
Simulated Groundwater ◽  
Long Term Behavior

Direct disposal of spent nuclear fuel (SNF) into the proposed unsaturated geologic repository at Yucca Mountain, NV is being studied at several laboratories, including Argonne National Laboratory. Corrosion tests with SNF are being conducted to understand the long-term behavior of SNF under conditions designed to simulate the unsaturated conditions at the site. The SNF used in this study was the Approved Testing Material (ATM)-106 with a bum-up of 43 MW·d/kg U. A sample of ATM-106 fuel was exposed to dripping simulated groundwater for 271 days; after this time the experiment was terminated and the material removed for further study. Details of the testing methodology have been given by Finn et al.,.Previous attempts to study SNF with TEM have used ion milled samples, in this study we prepared the samples by ultramicrotomy which reduced the radiological hazard substantially.

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

Mobilization/Retention of Radionuclides during Corrosion of High Burnup Spent Fuel and Backfill Materials in Salt Brines

2002 ◽  
Vol 757 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Near Field ◽  
Waste Form ◽  
Spent Fuel ◽  
The Matrix ◽  
Backfill Materials ◽  
Salt Brine ◽  
Solid Phases ◽  
Long Term Behavior ◽  
The Impact

ABSTRACTWith respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study the fuel alteration in contact with groundwater and near field materials. The aim of this work is to evaluate the impact of candidate backfill materials hydroxylapatite and magnetite on the overall corrosion behavior of this waste form in salt brine; both materials are used in corrosion tests together with spent fuel. The instant releases and the matrix dissolution rates appear to be similar in presence and in absence of any backfill material under study. However, Am,Np,Pu,U and Sr are retained at different ratios on the hydroxylapatite, on the magnetite and on the fuel sample, indicating possibly the formation of different radionuclide containing new solid phases.

The National Research Council, Study of the Isolation System for Geologic Disposal of Radioactive Wastes

1983 ◽  
Vol 26 ◽  
Author(s):  
Thomas H. Pigford
Keyword(s):  
Individual Dose ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Isolation System ◽  
Waste Forms ◽  
The Individual ◽  
Long Term Performance ◽  
Term Performance ◽  
Environmental Releases

ABSTRACTThis study was conducted for the U. S. Department of Energy by the Waste Isolation Systems Panel appointed by the National Academies of Science and Engineering. The panel was charged to review the alternative technologies available for Isolating of radioactive waste in mined geologic repositories, evaluate the performance benefits from these technologles as potential elements of a waste Isolation system, and identify appropriate technical criteria for satisfactory long-term performance of a geologic repository. Conceptual repositories in basalt, granite, salt, and tuff were considered. Site-specific data on geology, hydrology, and geochemical properties were evaluated and used to define parameters for estimating long-term environmental releases, supplemented when necessary by generic properties.The technology for solid waste forms and waste packages was reviewed and evaluated. Borosilicate glass and unreprocessed spent fuel are the waste forms appropriate for further testing and for repository designs. Testing in a simulated repository environment is necessary to develop an adeauate prediction of the long term performance of waste packages in a geologic repository. Back-up research and development on alternative waste forms should be continued. The expected functions of backfill placed between the rock and waste package need clearer definition and validation.The overall criterion to be used by federal agencies in designing a geologic waste-isolation system and in evaluating its nerformance has not yet been specified. As a guideline, the panel selected an average annual dose of 10-4 sieverts to a maximally exposed individual at any future time, if the exposure is from expected events such as the slow dissolution of waste solids in wet-rock repositories and the groundwater transport of dissolved radionuclides to the biosphere. Risks from unexpected events such as human intrusion were not evaluated.Calculations were made of the long-term isolation and environmental releases for conceptual repositories in basalt, granite, salt, and tuff. The major contributors to geologic isolation are the slow dissolution of key radioelements as limited by solubility and by diffusion and convection in groundwater surrounding the waste solids, long water travel times from the waste to the environment, and sorption retardation in the media surrounding the repository. Dilution by surface water can reduce the individual radiation exposures that can result from the small fraction of the waste radioactivity that may ultimately reach the environment. Estimates of environmental releases and individual doses were made both for unreprocessed spent fuel and for reprocessing wastes.Accelerated dissolution of waste exposed to groundwater during the period of repository heating was also considered. Long-term environmental releases of radioactivity from some repositories were calculated to cause doses to maximally exposed individuals that are several orders of magnitude below the Individual dose criterion of 10-4 Sieverts per year. Other conceptual repositories were found to not meet the individual dose criterion, although these repositories could still meet the radioactivity release limits in the standard proposed by the Environmental Protection Agency.The technology for geologic waste disposal has advanced to the state of a preliminary technical plan, suitable for testing, verification, and for pllot-facility confirmation. The waste Isolation program needs a reliable prediction of long-term performance that will serve as a basis for final design, construction, licensing, and waste emplacement.

scholarly journals Retention of Neptunium in Uranyl Alteration Phases Formed During Spent Fuel Corrosion

1997 ◽  
Vol 506 ◽  
Author(s):  
E. C. Buck ◽  
R. J. Finch ◽  
P. A. Finn ◽  
J. K. Bates
Keyword(s):  
Nuclear Fuel ◽  
Neutron Capture ◽  
Spent Nuclear Fuel ◽  
Secondary Phase ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Oxide Hydrate ◽  
First Time

ABSTRACTUranyl oxide hydrate phases are known to form during contact of oxide spent nuclear fuel with water under oxidizing conditions; however, less is known about the fate of fission and neutron capture products during this alteration. We describe, for the first time, evidence that neptunium can become incorporated into the uranyl secondary phase, dehydrated schoepite (UO3•0.8H2O). Based on the long-term durability of natural schoepite, the retention of neptunium in this alteration phase may be significant during spent fuel corrosion in an unsaturated geologic repository.

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