Alteration Behavior of High Burnup Spent Fuel in Salt Brine Under Hydrogen Overpressure and in Presence of Bromide

2006 ◽  
Vol 985 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler
Keyword(s):  
Dissolution Rate ◽  
Spent Fuel ◽  
Radiolysis Products ◽  
The Matrix ◽  
Matrix Bound ◽  
Salt Brine ◽  
High Burnup ◽  
Geological Formations ◽  
Α Activity

AbstractRecent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important ra-dionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bro-mide must be taken in consideration. Bromide is known to react with β/γ radiolysis products, thus counteracting the protective H2 effect. In the present experiments using high burnup spent fuel it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about10 in the presence of up to 10-3 M bromide and 3.2 bar H2 overpressure. However, concen-trations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bro-mide becomes less important, because the decrease of β/γ-activity results in a decrease of oxida-tive radicals, which react with bromide, while α activity will dominate the radiation field.

Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

2003 ◽  
Vol 807 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Dissolution Rate ◽  
Spent Fuel ◽  
Salt Solutions ◽  
Fuel Surface ◽  
Fuel Sample ◽  
Salt Brine ◽  
High Burnup ◽  
Interim Storage ◽  
The Impact

ABSTRACTDuring long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

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.

Congruent and Incongruent Radionuclide Release During Matrix Dissolution of Partly Oxidized High Burnup Spent Fuel

2000 ◽  
Vol 663 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
H. Geckeis
Keyword(s):  
Dissolution Rate ◽  
Masking Effect ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Ph Values ◽  
Fe Powder ◽  
The Matrix ◽  
Order Of Magnitude ◽  
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 in particular the UO2 matrix dissolution behavior and the associated release/retention of radionuclides in contact with groundwater. During long term fuel storage, fuel oxidation may not be avoided. Main issue of this work is to identify the impact on the corrosion of partly oxidized fuel of environmental conditions such as (1) the nature of solution contacting the matrix, the (2) presence/absence of CO2, (3) fixed pH values within a range between pH 7- pH 11, and (4) the presence/absence of corroding container material (Fe-powder). Dissolution tests with powdered oxidized spent fuel in various granite waters, and NaCl-brine resulted in matrix dissolution rates in the same order of magnitude for all investigated media (ca.5×10−4/d). The presence of CO2 and fixed pH values (pH 5 – 11) was without a distinct effect. The independence of the dissolution rate of the oxidized fuel matrix upon the nature of solution, pCO2, fixed pH values (5-11) can probably be explained by a masking effect of radiolysis. In presence of Fe powder the matrix dissolution rate was found to be slowed down by a factor of ca. 20, associated with strong retention effects of radionuclides.

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.

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 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.

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.

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

2005 ◽  
Vol 346 (1) ◽  
pp. 24-31 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Spent Fuel ◽  
Salt Brine ◽  
High Burnup ◽  
Radionuclide Release

Overview of the French research on the evolution of spent fuel rod after discharge from the reactor

2009 ◽  
Vol 1193 ◽  
Author(s):  
C. Ferry ◽  
C. Cappelaere ◽  
C. Jegou ◽  
J.P. Piron ◽  
M. Firon ◽  
...  
Keyword(s):  
Creep Behaviour ◽  
Creep Model ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Water Contact ◽  
Fuel Rod ◽  
Safety Margins ◽  
The Matrix ◽  
The One

AbstractSince 2006, French research on spent fuel has focused on the main issues related to transport and extended in-pool storage of spent fuel assembly. Studies on creep behaviour of irradiated cladding have resulted in a new creep model which is valid over a wide domain of temperature, internal pressure and time. Under nominal conditions, no evolution of the spent fuel rod is expected during in-pool storage. In case of defective fuel rods in the storage pool, the consequences of fuel alteration on the initial defect of the cladding depend on the matrix alteration rate and nature of the secondary phases formed. Considering the optional scenario of direct disposal, the long-term behaviour of the spent fuel is investigated focusing on helium consequences before water contact on the one hand and on the influence of repository conditions on matrix alteration on the other hand. The aim of the on-going studies is to improve the safety margins initially introduced in the radionuclide source term models.

Modelling experimental results on radiolytic processes at the spent fuel water interface. II. Radionuclides release

2006 ◽  
Vol 932 ◽  
Author(s):  
E. Cera ◽  
M. Grivé ◽  
J. Bruno ◽  
T.E. Eriksen
Keyword(s):  
Redox State ◽  
Kinetic Modelling ◽  
Initial Increase ◽  
Spent Fuel ◽  
Work Related ◽  
Radiolysis Products ◽  
The Matrix ◽  
Anoxic Environment ◽  
Bicarbonate Solutions

ABSTRACTExperimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2 surfaces in poising the redox state of solid/water systems as well as the radionuclides release behaviour. For this purpose, an experimental programme was developed consisting on dissolution experiments with PWR spent fuel fragments in an anoxic environment and by using different solution compositions.Some of the collected data has been previously published [1], specifically those data concerning radiolysis products and dissolution of the matrix. The results and the modelling tasks indicated an overall balance of the generated radiolytic species and that uranium dissolution was controlled by the oxidation of the spent fuel matrix in 10mM bicarbonate solutions while in the tests carried out at lower or without carbonate concentrations uranium in the aqueous phase was governed by the precipitation of schoepite.This paper is the continuation of a series accounting for the data and modelling work related to investigating the release behaviour of minor radionuclides from the spent fuel.Uranium concentrations as a function of time showed an initial increase until reaching a steady state, indicating a matrix dissolution control. The same behaviour is observed for neptunium, caesium, strontium, technetium and molybdenum indicating a congruent release of these elements with the major component of the fuel matrix. On the other hand, no cler tendency is observed for plutonium data where additional solubility limiting mechanisms may apply.Kinetic modelling of the trace elements: caesium, strontium, technetium and molybdenum is based on the congruent release of these elements with the major component of the fuel matrix. Rate constants have been determined. Kinetic modelling of neptunium data took also into account the subsequent precipitation as Np(IV) hydroxide. Finally, measured Pu concentrations may be explained by the precipitation of Pu(IV) and/or Pu(III) solid phases.

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