Kinetic Studies of Natural Uranium Minerals for the Long-Term Evolution of Spent Nuclear Fuel Under Oxidizing Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Ignasi Casas ◽  
E. Cera ◽  
J. Bruno
Keyword(s):  
Spent Nuclear Fuel ◽  
Kinetic Studies ◽  
Natural Uranium ◽  
Spent Fuel ◽  
Natural Systems ◽  
Term Evolution ◽  
Natural Analogues ◽  
Order Of Magnitude ◽  
Granitic Groundwater

ABSTRACTThe time scale of spent fuel dissolution studies is of the order of magnitude of 2 to 10 years, while the performance of a spent fuel repository should be assessed for much longer times (105-106 years). These time scales can be bridged using appropriate natural analogues. Among other important information, the study of natural systems can give insight of which can be the oxidative alteration of spent fuel in granitic environments. However, in studying such systems, thermodynamic and kinetic data of relevant natural solid phases are needed.In this work we present preliminary results of dissolution experiments carried out under oxidizing conditions with selected and well characterized natural samples of the alteration chain of uraninite (i.e., uraninite, schoepite, uranophane).The experiments have been performed using a synthetic granitic groundwater as a leachant, in contact with air and at 25 °C.

Review about the Effect of He on the Microstructure of Spent Nuclear Fuel in a Repository

MRS Advances ◽  
2016 ◽  
Vol 1 (62) ◽  
pp. 4147-4156 ◽  
Author(s):  
C. Ferry ◽  
J. Radwan ◽  
H. Palancher
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Concentration ◽  
Fuel Pellet ◽  
Spent Fuel ◽  
Operational Model ◽  
Matrix Microstructure ◽  
Natural Analogues ◽  
Order Of Magnitude ◽  
Radionuclide Release

ABSTRACTHelium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

The Assessment of the Long-Term Evolution of the Spent Nuclear Fuel Matrix by Kinetic, Thermodynamic and Spectroscopic Studies of Uranium Minerals.

1994 ◽  
Vol 353 ◽  
Author(s):  
Jordi Bruno ◽  
I. Casas ◽  
E Cera ◽  
R. C. Ewing ◽  
R. J. Finch ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spectroscopic Studies ◽  
Spent Fuel ◽  
Term Evolution ◽  
Uranium Minerals ◽  
The Stability

AbstractThe long term behaviour of spent nuclear fuel is discussed in the light of recent thermodynamic and kinetic data on mineralogical analogues related to the key phases in the oxidative alteration of uraninite. The implications for the safety assessment of a repository of the established oxidative alteration sequence of the spent fuel matrix are illustrated with Pagoda calculations. The application to the kinetic and thermodynamic data to source term calculations indicates that the appearance and duration of the U(VI) oxyhydroxide transient is critical for the stability of the fuel matrix.

Uranium-Series Ages of Secondary Uranium Minerals with Applications to the Long-Term Evolution of Spent Nuclear Fuel

1995 ◽  
Vol 412 ◽  
Author(s):  
R. J. Finch ◽  
J. Suksi ◽  
K. Rasilainen ◽  
R. C. Ewing
Keyword(s):  
Spent Nuclear Fuel ◽  
Natural Uranium ◽  
Uranium Deposits ◽  
Geologic Repository ◽  
Uranium Series ◽  
Term Evolution ◽  
Lower Solubility ◽  
Mineral Assemblages ◽  
Activity Ratios

AbstractUranium-series activity ratios for U(VI) minerals from the Shinkolobwe mine in southern Zaire indicate that these minerals have not experienced significant preferential loss of uranium since their formation more than 100,000 years ago. The minerals examined include rutherfordine, UO2CO3, schoepite, [(UO2)8O2(OH)12]·12H2O, becquerelite, Ca[(UO2)6O4(OH)6]·8H2O, and uranophane, Ca[(UO2)2(SiO3OH)2]·5H2O. No correlation between mineral species and mineral age was evident. The oxidative dissolution of primary uraninite (UO2+x) has maintained ground waters supersaturated with respect to all of the secondary U(VI) minerals, providing an inexhaustible source of dissolved U6+ for mineral formation and growth. As long as uraninite persists in an oxidizing environment, the assemblage of secondary U(VI) phases is determined by local ground water chemistry (including transitory changes), but not necessarily a unidirectional reaction path towards equilibrium with U(VI) minerals of lower solubility. Thus the Shinkolobwe mine displays a complex assemblage of U(VI) minerals that reflects variations in the availability of dissolved elements besides U. Similarly, for a geologic repository exposed to oxidizing waters, the assemblage of corrosion products that will form during the corrosion of spent UO2 fuel is likely to be as complex as mineral assemblages found in natural uranium deposits under similar conditions.

Specific outcomes of the research on the spent fuel long-term evolution in interim dry storage and deep geological disposal

2006 ◽  
Vol 352 (1-3) ◽  
pp. 246-253 ◽  
Author(s):  
C. Ferry ◽  
C. Poinssot ◽  
C. Cappelaere ◽  
L. Desgranges ◽  
C. Jegou ◽  
...  
Keyword(s):  
Long Term Evolution ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Dry Storage ◽  
Term Evolution ◽  
Deep Geological Disposal

Phase Relations of the Uranyl Oxide Hydrates and their Relevance to the Disposal of Spent Fuel

1990 ◽  
Vol 212 ◽  
Author(s):  
R. J. Finch ◽  
R. C. Ewing
Keyword(s):  
General Formula ◽  
Phase Relations ◽  
Molecular Water ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Structural Sheet ◽  
Natural Analogues ◽  
Ph Conditions ◽  
Image Position

ABSTRACTUranyl oxide hydrates, formed by the alteration of uraninite, are natural analogues for the long-term corrosion products of spent fuel in a geologic repository under oxidizing conditions. The uranyl oxide hydrates may be represented by the general formula:Pb-bearing hydrates require the addition of a neutral uranyl group into the structural sheet (UO2(OH)2) for each interlayer Pb ion. Distortion of the structure associated with the additional uranyl groups is reduced by replacing two structural hydroxyls with a structural oxygen and a molecular water. The general formula for the Pb-uranyl oxide hydrates is:This hypothesis explains the paragenetic sequences:1) schoepite ➛ billietite ➛ protasite ➛ bauranoite2) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ masuyite ➛ wölsendorfite3) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ ± masuyite ➛ sayrite ➛ curite, and indicates that, under relatively high pH conditions, schoepite will not be the long-term solubility-controlling phase for uranium in uranium-rich groundwaters.

Na/Ca selectivity coefficients of montmorillonite in perchlorate solution at different temperatures

2012 ◽  
Vol 1475 ◽  
Author(s):  
Aku Itälä ◽  
Arto Muurinen
Keyword(s):  
Cation Exchange ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Selectivity Coefficients ◽  
Crystalline Bedrock ◽  
Bentonite Buffer ◽  
Different Temperatures ◽  
C 50 ◽  
Exchange Selectivity

ABSTRACTThe Finnish spent nuclear fuel disposal is based on the Swedish KBS-3 concept in crystalline bedrock. The concept aims at long-term isolation and containment of spent fuel in copper canisters surrounded by bentonite buffer which mostly consists of montmorillonite. For the long-term modelling of the chemical processes in the buffer, the cation-exchange selectivity coefficients have to be known at different temperatures. In this work, the cation-exchange selectivity coefficients and cation-exchange isotherms were determined in batch experiments for montmorillonite at three different temperatures (25 °C, 50 °C and 75 °C). Five different ratios of NaClO4/Ca(ClO4)2 were used in the experimental solutions. After equilibration the solution and montmorillonite were separated and the solution analysed to get the desired exchange parameters. The experiments were modelled with a computational model capable of taking into account the physicochemical processes that take place in the experiment.

Behavior of UO2 in the RBMK-1000 Spent Fuel under Oxidizing Conditions

2004 ◽  
Vol 824 ◽  
Author(s):  
A. B. Kolyadin ◽  
V. Ya. Mishin ◽  
K. Ya. Mishin ◽  
A. S. Aloy ◽  
T. I. Koltsova
Keyword(s):  
Thermal Oxidation ◽  
Spent Nuclear Fuel ◽  
Oxidation Process ◽  
Spent Fuel ◽  
X Ray Diffraction ◽  
X Ray ◽  
Dry Storage ◽  
Different Temperatures ◽  
Xrd Techniques

AbstractThe oxidation of UO2–type spent nuclear fuel (SNF) in gaseousmedia was studied at different temperatures and oxygen contents using gravimetric and powder X-ray diffraction (XRD) techniques. The aim of the study was to determine the mechanism(s) of thermal-oxidation alteration of SNF during long-term dry storage. The samples used in the experiments were chips of RBMK-1000 fuel rods.Oxidation of UO2with a mean burn-up of 10.7 and 19.73 MW d/kg in humid air was observed at a temperature as low as 150°C. At 200°C nearly all of the UO2was transformed into U3O8 between 3500-4000 hours. In a humid nitrogen environment containing of 0.05-1.3 vol. % oxygen at 300°C, the UO2 completely transformed to U3O8 between 2500-3000 hours. Oxidation of UO2in samples with small amounts of jacket damage (e.g., <0.04 MM2)ll progresses more slowly and after â3000 hours the oxygen-to-uranium ratio was 2.56.Stabilization of the oxidation process was not observed in the fuel samples upto an O/U ratio of 2.4, which may be attributed to the smallburn-up of the fuel under investigation.

scholarly journals New trends in the reprocessing of spent nuclear fuel. Separation of minor actinides by solvent extraction

2016 ◽  
Vol 71 (1) ◽  
pp. 123
Author(s):  
Jerzy Narbutt
Keyword(s):  
Solvent Extraction ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Natural Uranium ◽  
Heterocyclic Ligands ◽  
High Level Nuclear Waste ◽  
Extraction Processes ◽  
High Level

<p>Recycling of actinides from spent nuclear fuel by their selective separation followed by transmutation in fast reactors will optimize the use of natural uranium resources and minimize the long-term hazard from high-level nuclear waste. This paper describes solvent extraction processes recently developed, aimed at the separation of americium from lanthanide fission products as well as from curium present in the waste. Depicted are novel poly-N-heterocyclic ligands used as selective extractants of actinide ions from nitric acid solutions or as actinide-selective hydrophilic stripping agents.</p>

scholarly journals A new approach to the recycling of spent nuclear fuel in thermal reactors within the REMIX concept

2020 ◽  
Vol 6 (2) ◽  
pp. 93-98
Author(s):  
Nikita V. Kovalev ◽  
Boris Ya. Zilberman ◽  
Nikolay D. Goletsky ◽  
Andrey B. Sinyukhin
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Natural Uranium ◽  
Spent Fuel ◽  
High Concentration ◽  
Energy Potential ◽  
New Approach ◽  
Reprocessed Uranium

A review of simulated nuclear fuel cycles with mixed uranium-plutonium fuel (REMIX) was carried out. The concept of REMIX fuel is one of the options for closing the nuclear fuel cycle (NFC), which makes it possible to recycle uranium and plutonium in VVER-1000/1200 thermal reactors at a 100% core loading. The authors propose a new approach to the recycling of spent nuclear fuel (SNF) in thermal reactors. The approach implies a simplified fabrication of mixed fuel when plutonium is used in high concentration together with enriched natural uranium, while reprocessed uranium is supposed to be enriched and used separately. The share of standard enriched natural uranium fuel in this nuclear fuel cycle is more than 50%, the share of mixed natU+Pu fuel is 25%, the rest is fuel obtained from enriched reprocessed uranium. It is emphasized that the new approach has the maximum economic prospect and makes it possible to organize the fabrication of this fuel and nuclear material cross-cycling at the facilities available in the Russian Federation in the short term. This NFC option eliminates the accumulation of SNF in the form of spent fuel assemblies (SFA). SNF is always reprocessed with the aim of further using the primary reprocessed uranium and plutonium. Non-recyclable in thermal reactors, burnt, reprocessed uranium, the energy potential of which is comparable to natural uranium, as well as secondary plutonium intended for further use in fast reactors, are sent as reprocessing by-products to the storage area.

Simfuel Dissolution Studies in Granitic Groundwater at 25°C

1990 ◽  
Vol 212 ◽  
Author(s):  
Amaia Sandino ◽  
I. Casas ◽  
K. Ollila ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Minor Component ◽  
Minor Components ◽  
Spent Fuel ◽  
Cumulative Release ◽  
Dissolution Studies ◽  
Leaching Experiments ◽  
Granitic Groundwater ◽  
Chemical Analogue

ABSTRACTThe dissolution of a non-radioactive chemical analogue of spent nuclear fuel (SIMFUEL) has been studied as a function of two different synthetic groundwaters at 25°C. The cumulative release of U, Mo, Ba, Y and Sr is presented after 170 days of total leaching time.The results obtained show the potential usefulness of SIMFUEL to ascertain the kinetics and mechanisms of dissolution of the minor components of spent fuel. In the case of Sr, a good correlation is found with the dissolution of this minor component measured in spent fuel leaching experiments.

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