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Long-term safety of radioactive waste disposal: Reaction of high burnup spent fuel and UO{sub 2} in saline brines at room temperature

Technical Report:

Abstract

In order to determine the long-term performance of spent fuel during direct disposal, high burnup fuel (50 MWd/kg U) has been exposed for 500 days to non-buffered saline solutions and to deionized water under static, anaerobic conditions at 25 C. In the absence of saturation effects (large solution volumes) a dissolution rate of 0,3% p.a. has been measured. Under more realistic conditions of large sample surface areas and small solution volumes, reaction rates were found to be orders of magnitude lower. The pH values remained neutral to slightly alkaline. Radiolysis did not lead to an oxidation of the solution. The releases of Sr90, Tc99, Np237 and Sb125 were controlled by the rate of dissolution of the fuel matrix, whereas the solution concentrations of U, Pu, Am and REE were governed by sorption and solubility effects. In the presence of iron, radiolytically produced oxygen has been consumed by iron corrosion and sorption on iron corrosion products effectively reduced the solution concentrations of most radionuclides. Studies of the corrosion of unirradiated UO{sub 2} have shown strong indications for the formation of U(IV) hydroxide. Under oxidizing conditions, only in few cases solubility limits were reached. In the determination of UO{sub 2} corrosion rates  More>>
Publication Date:
Aug 01, 1994
Product Type:
Technical Report
Report Number:
KFK-5377
Reference Number:
SCA: 360205; 052002; PA: DEN-95:0F0315; EDB-95:023075; SN: 95001321142
Resource Relation:
Other Information: PBD: Aug 1994
Subject:
36 MATERIALS SCIENCE; 12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; SPENT FUELS; CORROSION; URANIUM DIOXIDE; UNDERGROUND DISPOSAL; SALT CAVERNS; SAFETY; BURNUP; SIMULATION; CHEMICAL REACTIONS; BRINES; RADIOLYSIS; PH VALUE; DISSOLUTION; FISSION PRODUCT RELEASE; 360205; 052002; CORROSION AND EROSION; WASTE DISPOSAL AND STORAGE
OSTI ID:
10110205
Research Organizations:
Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Nukleare Entsorgungstechnik; Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Heisse Chemie
Country of Origin:
Germany
Language:
English
Other Identifying Numbers:
Journal ID: ISSN 0303-4003; Other: ON: DE95738047; CNN: Contract CEC F12W-0055; TRN: DE95F0315
Availability:
OSTI; NTIS (US Sales Only); INIS
Submitting Site:
DEN
Size:
58 p.
Announcement Date:
Jun 30, 2005

Technical Report:

Citation Formats

Grambow, B, Loida, A, Dressler, P, Geckeis, H, Diaz, P, Gago, J, Casas, I, Pablo, J de, Gimenez, J, and Torrero, M E. Long-term safety of radioactive waste disposal: Reaction of high burnup spent fuel and UO{sub 2} in saline brines at room temperature. Germany: N. p., 1994. Web.
Grambow, B, Loida, A, Dressler, P, Geckeis, H, Diaz, P, Gago, J, Casas, I, Pablo, J de, Gimenez, J, & Torrero, M E. Long-term safety of radioactive waste disposal: Reaction of high burnup spent fuel and UO{sub 2} in saline brines at room temperature. Germany.
Grambow, B, Loida, A, Dressler, P, Geckeis, H, Diaz, P, Gago, J, Casas, I, Pablo, J de, Gimenez, J, and Torrero, M E. 1994. "Long-term safety of radioactive waste disposal: Reaction of high burnup spent fuel and UO{sub 2} in saline brines at room temperature." Germany.
@misc{etde_10110205,
title = {Long-term safety of radioactive waste disposal: Reaction of high burnup spent fuel and UO{sub 2} in saline brines at room temperature}
author = {Grambow, B, Loida, A, Dressler, P, Geckeis, H, Diaz, P, Gago, J, Casas, I, Pablo, J de, Gimenez, J, and Torrero, M E}
abstractNote = {In order to determine the long-term performance of spent fuel during direct disposal, high burnup fuel (50 MWd/kg U) has been exposed for 500 days to non-buffered saline solutions and to deionized water under static, anaerobic conditions at 25 C. In the absence of saturation effects (large solution volumes) a dissolution rate of 0,3% p.a. has been measured. Under more realistic conditions of large sample surface areas and small solution volumes, reaction rates were found to be orders of magnitude lower. The pH values remained neutral to slightly alkaline. Radiolysis did not lead to an oxidation of the solution. The releases of Sr90, Tc99, Np237 and Sb125 were controlled by the rate of dissolution of the fuel matrix, whereas the solution concentrations of U, Pu, Am and REE were governed by sorption and solubility effects. In the presence of iron, radiolytically produced oxygen has been consumed by iron corrosion and sorption on iron corrosion products effectively reduced the solution concentrations of most radionuclides. Studies of the corrosion of unirradiated UO{sub 2} have shown strong indications for the formation of U(IV) hydroxide. Under oxidizing conditions, only in few cases solubility limits were reached. In the determination of UO{sub 2} corrosion rates one has to distinguish between the initial fast process of dissolving an oxidized surface layer of UO{sub 2,6} and the process of dissolution of the uranium dioxide beneath and the formation of a new layer of UO{sub 2,4}. Under anaerobic conditions reaction rates were significantly lower than under oxidizing conditions. (orig.) [Deutsch] Zur Bewertung der Langzeitstabilitaet im Hinblick auf die Direkte Endlagerung wurde hochabgebrannter Kernbrennstoff (50 MWd/kg U) unter statischen anaeroben Bedingungen in Kontakt gebracht mit ungepufferten Salzloesungen und mit reinem Wasser fuer Zeitraeume ueber 500 Tage bei 25 C. Unter Ausschluss von Saettigungseffekten (grosse Loesungsvolumina) wurde eine Aufloesungsrate von 0,3% pro Jahr gemessen. Unter realistischen Bedingungen grosser Brennstoffoberflaechen und kleiner Loesungsvolumina wurden um Groessenordnungen kleinere Reaktionsraten gemessen. Die pH-Werte in den Loesungen blieben bei fortschreitender Reaktion neutral bis leicht alkalisch. Radiolyse fuehrte nicht zu einer Oxidation der Loesung. Die Freisetzung von Sr90, Tc99, Np237 und Sb125 wird durch die Rate der Aufloesung der UO{sub 2}-Matrix limitiert, waehrend die Loesungskonzentrationen von U, Pu, Am und SEE durch Sorptions- und Loeslichkeitseffekte bestimmt werden. In Gegenwart von Eisen wird der gesamte, durch Radiolyse entstehende freie Sauerstoff bei der Korrosion des Eisens verbraucht und durch Sorption an Fe-Korrosionsprodukten reduzierten sich die Loesungskonzentrationen der untersuchten Nuklide erheblich. Korrosionsversuche mit unbestrahltem UO{sub 2} zeigen, dass sich unter reduzierenden Bedingungen die Bildung von U(IV)-Hydroxiden andeutet, waehrend unter oxidierenden Bedingungen nur in wenigen Faellen Loeslichkeitsgrenzen erreicht wurden. Bei der Ermittlung von Aufloesungsraten muss zwischen der schnellen Aufloesung einer oxidierten Oberflaechenschicht der Proben von einer Zusammensetzung UO{sub 2.66}, und der langsamen Aufloesung des unter der UO{sub 2.6} liegenden Urandioxids unter Bildung einer stationaeren Oberflaechenschicht mit der Zusammensetzung UO{sub 2.4} unterschieden werden. Unter anaeroben Bedingungen lagen die Reaktionsraten deutlich unter denen bei oxidierenden Bedingungen. (orig.)}
place = {Germany}
year = {1994}
month = {Aug}
}