Study of Actinides Incorporation in Thorium Phosphate-Diphosphate/Monazite Based Ceramics

2003 ◽  
Vol 802 ◽  
Author(s):  
Nicolas Clavier ◽  
Nicolas Dacheux ◽  
Renaud Podor ◽  
Philippe Le Coustumer
Keyword(s):  
Relative Density ◽  
Leaching Tests ◽  
Chemical Compatibility ◽  
Spent Fuel ◽  
Acidic Media ◽  
Dissolution Rates ◽  
Neutron Absorber ◽  
Pellet Form

ABSTRACTPhosphate materials are usually considered as potential candidates to perform the immobilization of actinides coming from an advanced reprocessing of spent fuel in the field of an underground repository. Among them, Thorium Phosphate-Diphosphate (TPD) and monazites have been already extensively studied. The elaboration of TPD/monazite based materials was thus envisaged in order to immobilize simultaneously tri- and tetravalent actinides and a neutron absorber. Two chemical ways of synthesis were considered and the compounds were easily prepared in the powder and in the pellet form. A good chemical compatibility was found between TPD and monazite since the properties of both phosphates were kept in these composites. Moreover, the relative density of the pellets reached 90 – 95 % of the value calculated from XRD data. The normalized dissolution rates determined in acidic media did not exceed 5.10−4 g.m−2.day−1 which confirmed the very good durability of such materials during leaching tests.

Qualification Requirements Design for Neutron Absorber Plate for Spent Fuel Racks in Domestic Nuclear Power Plants

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
You Shi ◽  
Dong Ning ◽  
Yi-zhong Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Aluminum Matrix ◽  
Practical Significance ◽  
Spent Fuel ◽  
Absorber Plate ◽  
Neutron Absorber ◽  
Fuel Storage ◽  
B4c Particle

Boron carbide (B4C) particle-reinforced aluminum matrix composite is the key material for use as neutron absorber plate in fuel storage applications for Generation III advanced passive nuclear power plants in China. This material has once depended upon importing with various restrictions so that it has meaningful practical significance to realize the localized manufacturing for this material in China. More importantly, since it is the first time for this material to be used in domestic plant, particular care should be taken to assure the formal supplied products exhibit high stabilized and reliable service in domestic nuclear engineering. This paper initiates and proposes a principle design framework from technical view in qualification requirements for this material so as to guide the practical engineering application. Aiming at neutron absorber materials supplied under practical manufacturing condition in engineering delivery, the qualification requirements define B4C content, matrix chemistry, 10B isotope, bulk density, 10B areal density, mechanical property, and microstructure as key criteria for material performance. The uniformity assessment as to different locations of this material is also required from at least three lots of material. Only qualified material meeting all of the qualification requirements should proceed to be verified by lifetime testing such as irradiation, corrosion, and thermal aging testing. Systematic and comprehensive performance assessments and verification for process stabilization could be achieved through the above qualification. The long-term service for this neutron absorber material in reliable and safe way could be convincingly expected in spent fuel storage application in China.

scholarly journals Development of a new neutron absorber material for dry casks and spent fuel pool racks

2016 ◽  
Vol 66 (3) ◽  
pp. 130-135
Author(s):  
Takutoshi Kondo ◽  
Xavier Clausse ◽  
Toshiaki Yamazaki ◽  
Hideki Hommo ◽  
Akiei Tanaka ◽  
...  
Keyword(s):  
Spent Fuel ◽  
Neutron Absorber ◽  
Absorber Material ◽  
Spent Fuel Pool

On the Role of H2 as an Inhibitor of UO2 Matrix Dissolution

2007 ◽  
Author(s):  
Juan Merino ◽  
Xavier Gaona ◽  
Lara Duro ◽  
Jordi Bruno ◽  
Aurora Marti´nez-Esparza
Keyword(s):  
Chemical Composition ◽  
Partial Pressure ◽  
Water Radiolysis ◽  
Water Interface ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Term Stability ◽  
Long Term Stability

The study of spent fuel behaviour under disposal conditions is usually based on conservative approaches assuming oxidising conditions produced by water radiolysis at the fuel/water interface. However, the presence of H2 from container corrosion can inhibit the dissolution of the UO2 matrix and enhance its long-term stability. Several studies have confirmed the decrease in dissolution rates when H2 is present in the system, although the exact mechanisms of interaction have not been fully established. This paper deals with a radiolytic modelling exercise to explore the consequences of the interaction of H2 with radicals generated by radiolysis in the homogeneous phase. The main conclusion is that in all the modelled cases the presence of H2 in the system leads to a decrease in matrix dissolution. The extent of the inhibition, and the threshold partial pressure for the inhibition to take place, both depend in a complex way on the chemical composition of the water and the type of radiation present in the system.

scholarly journals Comparison and Modeling of Aqueous Dissolution Rates of Various Uranium Oxides

1996 ◽  
Vol 465 ◽  
Author(s):  
S. A. Steward ◽  
E. T. Mones
Keyword(s):  
Dissolution Rate ◽  
Atmospheric Oxygen ◽  
Fold Increase ◽  
Stable Phase ◽  
Uranium Oxides ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Dissolution Rates ◽  
Intrinsic Dissolution ◽  
Carbonate Concentration

ABSTRACTThe purpose of this work has been to measure and model the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a geologic repository. When exposed to air at elevated temperature, spent fuel may form the stable phase U3O8. Dehydrated schoepite, UO3H2O, has been shown to exist in drip tests on spent fuel.Equivalent sets of U3O8 and UO3H2 dissolution experiments allowed a systematic examination of the effects of temperature (25–75°C), pH (8–10) and carbonate (2–200×10−4 molar) concentrations at atmospheric oxygen conditions.Results indicate that UO3H2O has a much higher dissolution rate (at least ten-fold) than U3O8 under the same conditions. The intrinsic dissolution rate of unirradiated U3O8 is about twice that of UO2. Dissolution of both U3O8 and UO3.H2O shows a very high sensitivity to carbonate concentration. Present results show a 25 to 50-fold increase in room-temperature UO3H2O dissolution rates between the highest and lowest carbonate concentrations.As with the UO2 dissolution data the classical observed chemical kinetic rate law was used to model the U3O8 dissolution rate data. The pH did not have much effect on the models, in agreement with the earlier analysis of the UO2 and spent fuel dissolution data,. However, carbonate concentration, not temperature, had the strongest effect on the U3O8 dissolution rate. The U3O8 dissolution activation energy was about 6000 cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and UO2 respectively.

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.

Uranium (iv) Dioxide and Simfuel as Chemical Analogues of Nuclear Spent Fuel Matrix Dissolution. A Comparison of Dissolution Results in a Standard Naci/NaHCO3 Solution

1994 ◽  
Vol 353 ◽  
Author(s):  
Jordi Bruno ◽  
I. Casas ◽  
E. Cera ◽  
J. de Pablo ◽  
J. GimÉnez ◽  
...  
Keyword(s):  
Dissolution Rate ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Dissolution Process ◽  
Experimental Comparison ◽  
Comparison Study ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Surface Sites ◽  
Radiolytic Oxidation

AbstractWe have carried out an experimental comparison study of the dissolution rates of unirradiated UO2 and SIMFUEL pellets and particles (100–300 μm) in a standard NaCI/NaHC03 solution, under oxidizing conditions. We have performed the experiments using batch and flow methodologies. Both methodologies gave similar results, indicating that the overall oxidation/dissolution process is the same in both cases. The results from the experiments indicate that under these conditions the dissolution process is both oxygen and bicarbonate promoted. The dissolution rates we obtained are: R=2.4 ± 0.8 mg U/m2 d for U02 and R= 0.17 ± 0.05 mg U/m2 d for SIMFUEL. The results of the experiments indicate that the dissolution rate under oxic conditions is clearly dependent on the number of U(VI) surface sites which for spent nuclear fuel is a function of the extent of radiolytic oxidation.

scholarly journals CRITICALITY SAFETY ANALYSIS OF GBC-32 SPENT FUEL CASK WITH IMPROVED NEUTRON ABSORBER CONCEPT

2021 ◽  
Vol 247 ◽  
pp. 17003
Author(s):  
Martin Lovecký ◽  
Jiří Závorka ◽  
Jana Jiřičková ◽  
Radek Škoda
Keyword(s):  
Fuel Assembly ◽  
Boron Content ◽  
High Efficiency ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Neutron Absorber ◽  
Safety Margins ◽  
Absorber Material ◽  
Inner Diameter ◽  
Criticality Safety

Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness.

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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