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.

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.

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.

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.

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.

A Comparison and Abstraction of the Commercial Spent Fuel Corrosion Experiments in the U.S.

2002 ◽  
Vol 757 ◽  
Author(s):  
Eric R. Siegmann
Keyword(s):  
Corrosion Rate ◽  
Spent Fuel ◽  
Release Rates ◽  
Water Contact ◽  
Contact Mode ◽  
Corrosion Rates ◽  
Single Temperature ◽  
Flow Through ◽  
Distribution Sensitivity ◽  
Temperature Dependent Model

ABSTRACTThis paper compares the results of three different fuel corrosion experiments by taking the existing corrosion data reported for various temperatures and recalculating corrosion rates at a single temperature of 25°C using a temperature dependent model developed elsewhere. Three types of light water reactor fuel corrosion tests (sometimes called dissolution or alteration tests) were performed in support of Yucca Mountain Project. The tests used three water contact modes and various fuel burnups. All measurements were adjusted for temperature differences and then compared. Five different isotopes (cesium, technetium, iodine, strontium, and, in the flow-through tests, uranium) were considered as a measure of corrosion. The data used represent over ten years of experiments with about nine different fuel types. Most experiments were with repository type fluids, containing dissolved constituents such as carbonate, calcium and silicon. The results show that all of the experiments predict similar fuel corrosion rates. Small differences in the isotope release rates are observed and incorporated in the abstracted uncertainty. Water contact mode (flow-through, batch, or drip) does not seem to be very important although the drip tests introduced larger variations. In developing a corrosion abstraction, all of the isotope measurements were considered equally. The distribution of release rates was used directly to develop the uncertainty. The mean corrosion rate was 1.8 × 10-4 fraction/year at 25°C (5%-95% range = 5.7 × 10-5 to 1.7 × 10-3). Using the derived corrosion rate for 25°C, and considering rapid axial splitting of the cladding, the CSNF fuel rod is expected to corrode in less than 2,000 years. The abstraction uses all the available experiments performed with water containing carbonates, silicon, or calcium and irradiated fuel to produce a corrosion rate distribution. Sensitivity studies using this corrosion rate abstraction in the TSPA-SR analysis show very small changes in dose (3%) in response to changes in the UO2 corrosion abstraction.

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.

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.

Radionuclides Release Model for Performance Assessment Studies of Spent Nuclear Fuel in Geological Disposal

2004 ◽  
Vol 824 ◽  
Author(s):  
Christophe Poinssot ◽  
Patrick Lovera ◽  
Cécile Ferry
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hydrogen Effect ◽  
Spent Fuel ◽  
Geological Disposal ◽  
The Matrix ◽  
Simple Kinetic Model ◽  
Scientific Results ◽  
Release Model

AbstractIn the framework of the research conducted on the long term evolution of spent nuclear fuel in geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of RN (Instant Release Fraction IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account all the scientific results currently available in the literature except the hydrogen effect. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated on the long term (rim, gap, grain boundaries). It allows to propose some reference bounding values for the IRF as a function of time of canister breaching and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the radiolytic species recombination and the influence of aqueous ligands and radiolytic oxidants were supposed to completely react with the fuel surface. Spent fuel performance was therefore demonstrated to deeply depend on the reactive surface area.

The IAEA Coordinated Research Programme on the Performance of High-Level Waste Forms and Packages Under Repository Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Vladimir S. Tsyplenkov
Keyword(s):  
Near Field ◽  
Research Programme ◽  
High Level Waste ◽  
Future Research ◽  
Spent Fuel ◽  
Waste Forms ◽  
Safety Assessments ◽  
High Level ◽  
Coordination Meeting

ABSTRACTThe IAEA initiated, in 1991, a Coordinated Research Programme (CRP), with the aim of promoting the exchange of information on the results obtained by different countries in the performance of high-level waste forms and waste packages under conditions relevant to final repository. These studies are being undertaken to obtain reliable data as input to safety assessments and environmental impact analyses, for final disposal purposes. The CRP includes studies on waste forms that are presently of interest worldwide: borosilicate glass, Synroc and spent fuel.Ten laboratories leading in investigation of high-level waste form performance have already joined the programme. The results of their studies and plans for future research were presented at the first Research Coordination Meeting, held in Karlsruhe, Germany, in November 1991. The technical contributions concentrated on effecting an understanding of dissolution mechanisms of waste forms under simulated repository conditions. A quantitative interpretation of the chemical processes in the near field is considered a prerequisite for long-term predictions and for the formulation of a "source term" for performance assessment studies.

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.

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