UO2 dissolution in high pH conditions of the Belgian Supercontainer

MRS Proceedings ◽

10.1557/opl.2012.589 ◽

2012 ◽

Vol 1475 ◽

Cited By ~ 3

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.

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Initial Results From Dissolution Testing of Spent Fuel Under Acidic Conditions

MRS Proceedings ◽

10.1557/proc-824-cc8.6 ◽

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.

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

MRS Advances ◽

10.1557/adv.2017.204 ◽

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.

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

MRS Proceedings ◽

10.1557/proc-932-85.1 ◽

2006 ◽

Vol 932 ◽

Cited By ~ 1

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.

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

MRS Proceedings ◽

10.1557/proc-932-77.1 ◽

2006 ◽

Vol 932 ◽

Cited By ~ 3

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.

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

MRS Proceedings ◽

10.1557/proc-212-241 ◽

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.

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On the Role of H2 as an Inhibitor of UO2 Matrix Dissolution

11th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B ◽

10.1115/icem2007-7109 ◽

2007 ◽

Author(s):

Juan Merino ◽

Xavier Gaona ◽

Lara Duro ◽

Jordi Bruno ◽

Aurora Marti´nez-Esparza

Keyword(s):

Chemical Composition ◽

Partial Pressure ◽

Water Radiolysis ◽

Water Interface ◽

Spent Fuel ◽

Dissolution Rates ◽

Term Stability ◽

Long Term Stability

The study of spent fuel behaviour under disposal conditions is usually based on conservative approaches assuming oxidising conditions produced by water radiolysis at the fuel/water interface. However, the presence of H2 from container corrosion can inhibit the dissolution of the UO2 matrix and enhance its long-term stability. Several studies have confirmed the decrease in dissolution rates when H2 is present in the system, although the exact mechanisms of interaction have not been fully established. This paper deals with a radiolytic modelling exercise to explore the consequences of the interaction of H2 with radicals generated by radiolysis in the homogeneous phase. The main conclusion is that in all the modelled cases the presence of H2 in the system leads to a decrease in matrix dissolution. The extent of the inhibition, and the threshold partial pressure for the inhibition to take place, both depend in a complex way on the chemical composition of the water and the type of radiation present in the system.

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The Effect of Nearfield Constraints on the Corrosion Behavior of High Burnup Spent Fuel

MRS Proceedings ◽

10.1557/proc-932-72.1 ◽

2006 ◽

Vol 932 ◽

Cited By ~ 3

Author(s):

Andreas Loida ◽

Manfred Kelm ◽

Bernhard Kienzler ◽

Horst Geckeis ◽

Andreas Bauer

Keyword(s):

Host Rock ◽

Near Field ◽

Experimental Studies ◽

Spent Fuel ◽

Mutual Support ◽

Significant Extent ◽

Dissolution Rates ◽

The Matrix ◽

Container Material

ABSTRACTThe long-term immobilization for individual radioelements released from the waste form “spent fuel” in solid phases upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. Related experimental studies comprise effects induced by the presence of Fe based container material, and near field materials other than Fe for a rock salt environment. The effect of the presence of an argillaceous host rock containing organic matter and pyrite on fuel alteration was studied in addition. The results have shown that oxidative radio-lysis products were found to be consumed at a significant extent by the metallic Fe and by the argillaceous host rock. Under these conditions a decrease at a factor of ca.100 for both the matrix dissolution rates and the solution concentrations of U and Pu was found. There is mutual support between the matrix dissolution rates, the solution concentrations and the amounts of oxygen encountered during the experiments under various conditions controlled by the presence of near field materials under study.

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Dissolution Rates of DWPF Glasses from Long-Term PCT

MRS Proceedings ◽

10.1557/proc-465-149 ◽

1996 ◽

Vol 465 ◽

Cited By ~ 6

Author(s):

W. L. Ebert ◽

S.-W. Tam

Keyword(s):

Dissolution Rate ◽

Surface Area ◽

Spherical Particles ◽

Dissolution Rates ◽

Glass Dissolution ◽

Dissolution Tests ◽

Dissolution Model ◽

Advanced Stages ◽

Radionuclide Release

ABSTRACTWe have characterized the corrosion behavior of several Defense Waste Processing Facility (DWPF) reference waste glasses by conducting static dissolution tests with crushed glasses. Glass dissolution rates were calculated from measured B concentrations in tests conducted for up to five years. The dissolution rates of all glasses increased significantly after certain alteration phases precipitated. Calculation of the dissolution rates was complicated by the decrease in the available surface area as the glass dissolves. We took the loss of surface area into account by modeling the particles to be spheres, then extracting from the short-term test results the dissolution rate corresponding to a linear decrease in the radius of spherical particles. The measured extent of dissolution in tests conducted for longer times was less than predicted with this linear dissolution model. This indicates that advanced stages of corrosion are affected by another process besides dissolution, which we believe to be associated with a decrease in the precipitation rate of the alteration phases. These results show that the dissolution rate measured soon after the formation of certain alteration phases provides an upper limit for the long-term dissolution rate, and can be used to determine a bounding value for the source term for radionuclide release from waste glasses. The long-term dissolution rates measured in tests at 20,000 m−1 at 90°C in tuff groundwater at pH values near 12 are about 0.2,0.07, and 0.04 g/(m2•d) for the Environmental Assessment glass and glasses made with SRL 131 and SRL 202 frits, respectively.

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Rational Design of Aptamer Switches with Programmable pH Response

10.1101/2020.03.14.992040 ◽

2020 ◽

Author(s):

Ian A.P. Thompson ◽

Liwei Zheng ◽

Michael Eisenstein ◽

H. Tom Soh

Keyword(s):

Rational Design ◽

Molecular Devices ◽

Design Strategies ◽

Ph Response ◽

Dna Motif ◽

Alkaline Conditions ◽

Previous Design ◽

Ph Conditions ◽

Target Binding ◽

Ph Range

AbstractAptamer switches that respond sensitively to local variations in pH could enable the development of molecular devices that greatly enhance diagnostic accuracy and therapeutic efficacy. Previous design strategies relied on inserting a known pH-sensitive DNA motif into the aptamer structure. Unfortunately, their performance was ultimately governed by the intrinsic pH-response of the inserted motif and could not be readily tuned to operate across arbitrary pH ranges. In this work, we present a general methodology for converting virtually any aptamer into a pH-responsive switch that can be modulated to undergo a strong change in affinity - in acidic, neutral, or alkaline conditions. This was achieved by inserting two orthogonal motifs that can be manipulated in parallel to tune sensitivity to different pH conditions without altering the core sequence of the aptamer itself. Using a single ATP aptamer, we achieve pH-dependent target binding and release under diverse pH conditions, producing as much as 1,000-fold pH-induced change in affinity. Importantly, we also demonstrate the novel capability to predictably engineer aptamer switches that only exhibit strong target affinity within a narrowly defined pH range. Our approach should offer a highly generalizable strategy for integrating pH-responsiveness into molecular devices.

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