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

2006 ◽  
Vol 932 ◽  
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.

Spent Fuel Corrosion Behavior in Salt Solution in the Presence of Hydrogen Overpressure

2001 ◽  
Author(s):  
Andreas Loida ◽  
Bernd Grambow ◽  
Horst Geckeis
Keyword(s):  
Gas Phase ◽  
Spent Fuel ◽  
Nacl Solution ◽  
Release Rates ◽  
Corrosion Rates ◽  
The Matrix ◽  
Matrix Bound ◽  
Safety Assessments ◽  
Container Material

Abstract The simultaneous corrosion of spent fuel and Fe-based container material is characterized by the formation of large amounts of hydrogen, which control the composition of the gas phase. Various experimental data indicate that the matrix dissolution rate and the release rates of important radionuclides decrease, if the H2 overpressure increases. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution rates, and to take credit from the effect of hydrogen overpressure in long-term safety assessments of the repository, a detailed experimental investigation has been initiated. High burnup spent fuel is being corroded under anoxic conditions in the absence of carbonate in 5m NaCl solution under an external H2 overpressure of 3.3 bar. This pressure is in the same range as observed in a long-term test using spent fuel and Fe-powder. Results obtained after 117 days of testing show that due to constant or decreasing concentrations of Sr and other matrix bound radionuclides, corrosion rates were not measurable indicating a stop of matrix dissolution or very low long-term rates. Grain boundary release of Cs and fission gases was found to continue under hydrogen overpressure. Compared to tests in the absence of hydrogen solution concentrations decreased by about ca. 1.5 orders of magnitude for U (10−8 M), Am, Eu (10−10 M), whereas the decrease of Np (3×10−10 M), Tc (5×10−9 M) and Pu (4×10−9 M) concentrations was found to be less significant.

Study of the Consequences of Secondary Water Radiolysis within and Surrounding a Defective Canister

2000 ◽  
Vol 663 ◽  
Author(s):  
Jinsong Liu ◽  
Bo Strömberg ◽  
Ivars Neretnieks
Keyword(s):  
Near Field ◽  
Experimental Studies ◽  
Solubility Limit ◽  
Balance Model ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Secondary Minerals ◽  
First Case ◽  
Bentonite Buffer

ABSTRACTA model has been developed to study the effects of secondary water radiolysis caused by dispersed radionuclides in a bentonite buffer surrounding a copper canister. The secondary radiolysis is the radiolysis caused by radionuclides that have been released from the spent fuel and are present either as solutes in the pore-water, as sorbed species on the surface of other minerals, or as secondary minerals. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The dissolution of the spent fuel is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases have been considered with the purpose to illustrate the behaviors of both conservative and non-conservative nuclides. The nuclides that are most relevant are those expected to be the dominant α-emitters in the long-term (e.g. 239Pu, 240Pu, 241Am). In the first case it is assumed that there is no precipitation of secondary minerals of the relevant radionuclides inside the canister. In the second case it is assumed that the radionuclide concentration within the canister is controlled by its respective solubility limit. The radionuclide released to the surrounding buffer is then predicted using a mass balance model. The modelling results show that in both cases, the spent fuel will not be oxidized at a rate significantly faster compared to the case where secondary radiolysis is completely neglected. In the first case, however, a large domain of the near-field can be oxidized due to a much faster depletion of reducing minerals in the buffer, compared to the case where secondary radiolysis is neglected. In the second case, the effects of the secondary water radiolysis will be quite limited.

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.

Measurement of Soluble Nuclide Dissolution Rates from Spent Fuel

1989 ◽  
Vol 176 ◽  
Author(s):  
C. N. Wilson ◽  
W. J. Gray
Keyword(s):  
Yucca Mountain ◽  
Water Soluble ◽  
Spent Fuel ◽  
Release Rates ◽  
Dissolution Rates ◽  
Advantages And Disadvantages ◽  
Water Composition ◽  
The Matrix ◽  
Flow Through

ABSTRACTGaining a better understanding of the potential release behavior of water-soluble radionuclides is the focus of new laboratory spent fuel dissolution studies being planned in support of the Yucca Mountain Project. Previous studies have suggested that maximum release rates for actinide nuclides, which account for most of the long-term radioactivity in spent fuel, should be solubility-limited and should not depend on the characteristics or durability of the spent fuel waste form. Maximum actinide concentrations should be sufficiently low to meet the NRC annual release limits. Potential release rates for soluble nuclides such as 99Tc, 135Cs, 14C and 129I, which account for about 1-2% of the activity in spent fuel at 1000 years, are less certain and may depend on processes such as oxidation of the fuel in the repository air environment.Dissolution rates for several soluble nuclides have been measured from spent fuel specimens using static and semi-static methods. However, such tests do not provide a direct measurement of fuel matrix dissolution rates that may ultimately control soluble-nuclide release rates. Flow-through tests are being developed as a potential supplemental method for determining the matrix component of soluble-nuclide dissolution. Advantages and disadvantages of both semi-static and flow-through methods are discussed. Tests with fuel specimens representing a range of potential fuel states that may occur in the repository, including oxidized fuel, are proposed. Preliminary results from flow-through tests with unirradiated UO2 suggesting that matrix dissolution rates are very sensitive to water composition are also presented.

Constraints by Experimental Data for Modeling of Radionuclide Release from Spent Fuel

1989 ◽  
Vol 176 ◽  
Author(s):  
Bernd Grambow ◽  
L.O. Werme ◽  
R.S. Forsyth ◽  
J. Bruno
Keyword(s):  
Experimental Data ◽  
Product Layer ◽  
Spent Fuel ◽  
Upper Limits ◽  
The Matrix ◽  
The Usa ◽  
Oxidant Production ◽  
Release Data ◽  
Radionuclide Release

ABSTRACTComparison of spent fuel corrosion data from nuclear waste management projects in Canada, Sweden and the USA strongly suggests that the release of 90Sr to the leachant can be used as a measure of the degradation (oxidation/dissolution) of the fuel matrix. A surprisingly quantitative similarity in the 90 Sr release data for fuel of various types (BWR, PWR, Candu), linear power ratings and burnups leached under oxic conditions was observed in the comparison. After 1000 days of leachant contact, static or sequential, the fractional release rates for 90Sr (and for cesium nuclides) were of the order of 10−7/d.The rate of spent fuel degradation (alteration) under oxic conditions can be considered to be controlled either by the growth rates of secondary alteration products, by oxygen diffusion through a product layer, by the rate of formation of radiolytic oxidants or by solubility-controlled dissolution of the matrix. These processes are discussed. Methods for determining upper limits for long-term 90Sr release, and hence fuel degradation, have been derived from the experimental data and consideration of radiolytic oxidant production.

The geological research in France - The Dosssier 2005 Argile

2006 ◽  
Vol 985 ◽  
Author(s):  
Frederic Plas ◽  
Jacques WENDLING
Keyword(s):  
Near Field ◽  
Concept Design ◽  
Dose Calculations ◽  
Spent Fuel ◽  
Geological Medium ◽  
Source Data ◽  
Underground Research Laboratory ◽  
Broad Outline ◽  
Geological Research

AbstractAt the end of fifteen years of researchs defined by the French act of December 30, 1991 on radwaste management, Andra gave a report, “Dossier Argile 2005”, which concluded with the feasibility of a reversible disposal in the argillaceous Callovo-Oxfordien formation studied by means of an underground research laboratory at Meuse/Haute-Marne site. Starting from source data like the characteristics of the geological medium and the waste inventory, the process followed by Andra to achieve at this conclusion is of type sequential and iterative between concept design, scientific knowledge, in particular that of the phenomenological evolution of the reposiroty and its geological environment from operating period to long term, and Safety assessment. The “Dossier Argile 2005” covers a broad radwaste inventory, ILLW, HLW and Spent Fuel, so that it makes it possible to cover whole of the technological, scientific and safety topics. This article will give an overview of the geological disposal studies in France and draw the main conclusion of the Dossier 2005 Argile. It will be focused on the near field (Engineering components and near field host rock), while considering if necessary its integration within the whole system. After a short description of the concepts (incl. waste inventory and the characteristics of the Meuse/Haute the Marne site) and the functions of the components of repository and geological medium, one will describe successively the broad outline of the phenomenological evolution of repository and the geological medium in near field, by in particular releasing the time scales of processes and uncertainties of knowledge. On this basis, one will indicate the safety scenarios which were considered and the broad outline of performance and dose calculations. Lessons learn from the Dossier 2005 Argile will be discussed and perspective and priority for future will be indicated.

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

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.

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.

Pitting Studies Under Anoxic Conditions on Candidate Container Materials AISI 316L hMo and UHB 904L for The Disposal of HLW in Argillaceous Formations

2003 ◽  
Vol 807 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts
Keyword(s):  
Stainless Steel ◽  
Stainless Steels ◽  
Near Field ◽  
Austenitic Stainless Steels ◽  
Aisi 316L ◽  
Progressive Change ◽  
Anoxic Conditions ◽  
The Impact ◽  
Container Material

ABSTRACTStainless steel is being envisaged as the primary candidate container material for the final disposal of vitrified HLW in deep geological argillaceous formations in Belgium. The impact of an evolving underground repository environment, i.e. a progressive change from oxic to anoxic conditions (due to the consumption of entrapped oxygen), on the pitting behaviour of austenitic stainless steels AISI 316L hMo and UHB 904L was studied. CPP-experiments were performed in synthetic solutions, which are representative for the near-field chemistry of an underground repository. The solutions contained various amounts of Cl- (100–50,000 mg/L) at near-neutral pH. Experiments were conducted at 16 and 90°C.AISI 316L hMo and UHB 904L will not be subjected to immediate pitting problems neither under oxic, nor under anoxic conditions. However, AISI 316L hMo could present long-term pitting problems under oxic conditions. Pits are much easier initiated on AISI 316L hMo, for both oxic and anoxic conditions. The pits propagate in a rather similar manner under oxic conditions for both alloys, whereas under anoxic conditions the pits formed on AISI 316L hMo are much deeper. AISI 316L hMo is more susceptible to crevice attack.

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.

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