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.

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.

Modelling of the Waste Form Behaviour in a Wet Near-Surface Repository Site Over Extended Time Periods

2001 ◽  
Author(s):  
Michael I. Ojovan ◽  
Natalia V. Ojóvan ◽  
Irene V. Startceva ◽  
Zoja I. Golubeva ◽  
Alexander S. Barinov
Keyword(s):  
Experimental Data ◽  
Surface Area ◽  
Salt Content ◽  
Volume Ratio ◽  
Waste Glass ◽  
Near Surface ◽  
Waste Forms ◽  
Time Periods ◽  
Radionuclide Release

Abstract A mathematical model was used to predict radionuclide release from bitumen and glass waste forms over extended time periods. To calculate some model parameters, we used experimental data derived from 12yr field tests with six borosilicate waste glass blocks (each ∼30 kg in weight) and a bitumen block (310 kg), containing real intermediate-level NPP operational waste (NaNO3, 86 wt.% of a dry salt content; 137Cs, 82% of the radioactive inventory). Specific radioactivities of the glass material containing 35 wt.% waste oxides were βtot(90Sr+90Y), 3.74×106 Bq/kg, and αtot(239Pu), 1.3×104Bq/kg. The bitumen block with ∼31 wt.% salt content and βtot(90Sr+90Y), 4.0·106 Bq/kg, and αtot(239Pu), 3.0×103 Bq/kg was manufactured on base of a hard bitumen BN-IV. Tests with the waste forms were performed under saturated conditions of an experimental near-surface repository with a free access of groundwater to the waste blocks through a covering of host loamy soil and backfill of coarse sand. The way used to quantify the amount of leached radioactivity was to measure the volume and radioactivity concentrations of contacting groundwater. In the model, radionuclide release from the waste glass is assumed to be controlled by the processes of diffusion limited ion exchange and glass network dissolution. The mechanism of radionuclide release from the bitumen matrix is believed to remain the same throughout the long-term storage period, except for the initial stage when an enhanced leaching from the surface layer occurs. This long-term release is assumed to be controlled by diffusion of radionuclides through the bitumen matrix. So, identical formulae were applied to calculate the values of leached radioactivity fractions for two waste forms. Radioactivity release curves were plotted for field data and calculation results. For both waste forms, there was good agreement between the modelled and available experimental data. According to the modelling results, fmax = 2.3×10−3% of the initial radioactivity will release from the waste glass into the environment within a proposed institutional control period of 300 years under conditions of the near-surface repository and in the absence of additional engineered barriers. For the bitumen block and the same 300-yr period, the total (maximum) leached radioactivity fraction will be fmax = 4.2×10−3%. The main result of the modelling and experimental studies concerning the leaching behaviour of the bituminised and vitrified waste materials is that the fractional radioactivity release for two waste forms is on the same order of magnitude. Numerical release values per a unit of a surface area to volume ratio are also rather close for two waste forms (exposed surface area to volume ratio for the bitumen block is 2 to 4 times greater then for the glass).

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.

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.

Processes Controlling Radionuclide Release from Spent Fuel

1994 ◽  
Vol 353 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
H. Geckeis ◽  
P. Dressler
Keyword(s):  
Volume Ratio ◽  
Reaction Rates ◽  
Geometric Constraints ◽  
Spent Fuel ◽  
Oxidative Dissolution ◽  
Strong Decrease ◽  
Radiolysis Products ◽  
Upper Limits ◽  
Fuel Behavior ◽  
Radionuclide Release

AbstractDissolution of spent fuel has been studied in saline, anaerobe, carbonate free solutions. Processes controlling spent fuel dissolution and associated radionuclide release are radiolytically controlled oxidative dissolution, sorption on container, solubility and coprecipitation. Upper limits for oxidative dissolution rates are given by the production rates of oxidative radiolysis products. This limitation leads to a strong decrease in surface area normalized reaction rates with increasing surface to volume ratio (S/V) and imposes geometric constraints on prediction of spent fuel behavior in a repository. Solution concentrations of Am during spent fuel corrosion were about 5 orders of magnitude lower than the solubility of Am(OH)3(s) and are likely controlled by coprecipitation. Pu concentrations may be controlled by Pu(VI) or Pu(IV) (hydr)oxides.

Effect of Spent Fuel Burnup and Composition on Alteration of the U(Pu)O2 Matrix

2000 ◽  
Vol 663 ◽  
Author(s):  
C. Jégou ◽  
S. Peuget ◽  
J.F. Lucchini ◽  
C. Corbel ◽  
V. Broudic ◽  
...  
Keyword(s):  
Release Kinetics ◽  
Nuclear Waste Repository ◽  
Spent Fuel ◽  
Static Mode ◽  
Mox Fuel ◽  
The Matrix ◽  
Leaching Experiments ◽  
Experimental Findings ◽  
Radionuclide Release ◽  
Kinetics Of

ABSTRACTFor a potential performance assessment of direct disposal 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.Leaching experiments were performed with five spent fuel samples from French power reactors (four UO2 fuel samples with burnup ratings of 22, 37, 47 and 60 GWd·tHM−1 and a MOX fuel sample irradiated to 47 GWd·tHM−1) to determine the release kinetics of the matrix containing most (over 95%) of the radionuclides. The experiments were carried out with granitic groundwater on previously leached sections of clad fuel rods in static mode, in an aerated medium at room temperature (25°C) in a hot cell.After 1000 or 2000 days of leaching, the Sr/U congruence ratios for all the UO2 fuel samples ranged from 1 to 2; allowing for the experimental uncertainty, strontium can thus be considered as a satisfactory matrix alteration tracer. No significant burnup effect was observed on the alteration of the UO2 fuel matrix. The daily strontium release factor was approximately 1 à 10−7 d−1 for UO2 fuel, and five to six times higher for MOX fuel. Several alteration mechanisms (radiolysis, solubility, precipitation/clogging) are examined to account for the experimental findings.

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.

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