scholarly journals Results from Long-Term Dissolution Tests Using Oxidized Spent Fuel

1990 ◽  
Vol 212 ◽  
Author(s):  
Charles N. Wilson
Keyword(s):  
Test Site ◽  
Fused Silica ◽  
Spent Fuel ◽  
Test Cycle ◽  
Release Rates ◽  
Fuel Oxidation ◽  
Dissolution Tests ◽  
J 13 ◽  
Fractional Release

ABSTRACTTwo semi-static dissolution tests using oxidized PWR spent fuel specimens are being conducted under ambient hot cell conditions in Nevada Test Site J-13 well water and unsealed fused silica vessels. The test specimens were oxidized at 250°C in air to bulk oxygen-to-metal (O/M) values of 2.21 and 2.33. Following an initial 191-day test cycle, the specimens were restarted in fresh J-13 water for a second long-term test cycle. Results through the first 40 months of Cycle 2 are compared with results from similar tests at 25°C and 85°C using unoxidized spent fuel specimens.Increased concentrations of U, Am, Cm and Np were measured in 0.4- μm filtered samples from the oxidized fuel tests compared to the unoxidized fuel tested at 25°C; Pu concentrations were not affected by the fuel oxidation state. Most of the Am and Cm, and a portion of the Pu, measured in 0.4-μm filtered samples was removed by 2-nm filtration. Fission product release results were normalized to specimen inventories and reported as fractional release. No attempt was made to normalize the data to surface area. Initial 99Tc release was greatly increased, and prolonged increases in the fractional release rates of 99Tc and 129I occurred as a result of fuel oxidation. Fractional release rates for 137Cs and 90Sr from oxidized fuel eventually decreased to levels similar to those observed with unoxidized fuel after equivalent testing times, suggesting that matrix dissolution rates normalized to fuel mass were not increased as a result of oxidation.

Experimental study of the Dissolution of spent fuel at 85°C in Natural Ground Water

1986 ◽  
Vol 84 ◽  
Author(s):  
Charles N. Wilson ◽  
Henry F. Shaw
Keyword(s):  
Environmental Protection Agency ◽  
Fused Silica ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Waste Package ◽  
Fuel Rod ◽  
Dissolution Tests ◽  
J 13 ◽  
End Caps

AbstractSemi-static dissolution tests using pressurized water reactor spent fuel rod segments and NNWSI reference J-13 well water in sealed stainless steel vessels at 85°C are being conducted in support of the Waste Package Task of the NNWSI Project. Test specimens include: bare fuel plus the empty cladding hulls, fuel rod segments with artificially induced cladding defects and water-tight end caps, and undefected fuel rod segments with water-tight end caps. The test conditions approximate those expected in the proposed NNWSI Project repository when the waste package has cooled sufficiently to allow water to enter a breached container and contact the fuel rods, some of which may exhibit various degrees of cladding failure. Periodic solution samples (unfiltered and filtered) were analyzed for most radionuclides for which cumulative release limits are listed by the U.S. Environmental Protection Agency. Results from the first six-month cycle of the 85° C tests are presented and are compared with results from the first cycle of a previous test series run at 25° in fused silica test vessels.

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.

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.

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.

scholarly journals Derivation of a Waste Package Source Term for NNWSI from the Results of Laboratory Experiments

1985 ◽  
Vol 50 ◽  
Author(s):  
Virginia M. Oversby ◽  
Charles N. Wilson
Keyword(s):  
Laboratory Experiments ◽  
Source Term ◽  
Yucca Mountain ◽  
Maximum Amount ◽  
Well Water ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
J 13 ◽  
Fractional Release

AbstractResults are presented for the dissolution of Turkey Point pressurized water reactor (PWR) spent fuel in J-13 well water at ambient hot cell temperatures. These results are compared with those previously obtained on Turkey Point fuel in deionized water, on H. B. Robinson PWR fuel in J-13 water, and by other workers using various fuels in dilute bicarbonate groundwaters. A model is presented that represents the conditions under which maximum dissolution of spent fuel could occur in a repository sited at Yucca Mountain, Nevada. Using an experimentally determined upper limit of 5 mg/l for uranium solubility in J-13 water, a fractional release rate of 6.4 × 10−8 per year is obtained by assuming that all water entering the repository carries away the maximum amount of uranium.

Initial Results From Dissolution Testing of Spent Fuel Under Acidic Conditions

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.

scholarly journals Reexamining the Dissolution of Spent Fuel: A Comparison of Different Methods for Calculating Rates

2004 ◽  
Vol 824 ◽  
Author(s):  
Brady D. Hanson ◽  
Ray B. Stout
Keyword(s):  
Surface Area ◽  
Initial Surface ◽  
Spent Fuel ◽  
Release Rates ◽  
Dissolution Rates ◽  
Effective Surface ◽  
Past Data ◽  
Geometric Surface ◽  
Fractional Release ◽  
Geometric Surface Area

AbstractDissolution rates for spent fuel have typically been reported in terms of a rate normalized to the surface area of the specimen. Recent evidence has shown that neither the geometric surface area nor that measured with BET accurately predicts the effective surface area of spent fuel. Dissolution rates calculated from results obtained by flowthrough tests were reexamined comparing the cumulative releases and surface area normalized rates. While initial surface area is important for comparison of different rates, it appears that normalizing to the surface area introduces unnecessary uncertainty compared to using cumulative or fractional release rates. Discrepancies in past data analyses are mitigated using this alternative method.

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.

Long Term Modelling of Spent Fuel Oxidation/Dissolution Under Repository Conditions

2001 ◽  
Author(s):  
Juan Merino ◽  
Esther Cera ◽  
Jordi Bruno ◽  
Trygve Eriksen ◽  
Javier Quiñones ◽  
...  
Keyword(s):  
Source Term ◽  
Redox System ◽  
Water Interface ◽  
Spent Fuel ◽  
Fuel Oxidation ◽  
The Subject ◽  
The Stability ◽  
Thermodynamic Processes

Abstract A model to study the stability of the spent fuel under repository conditions has been developed. The fuel-water interface is a dynamic redox system, where oxidising conditions due to the radiolysis of water can lead to the release of the uranium and the radionuclides embedded in the fuel matrix. Both kinetic and thermodynamic processes have been taken into account. Special attention is given to the unit rate of matrix oxidation/dissolution, which has been the subject of a specific radiolytic model. The findings of this work have important implications for the applicability of solubility limits in establishing source term models.

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.

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