Spent Fuel Leaching under Anoxic Conditions and the Effect of Canister Materials

2003 ◽  
Vol 807 ◽  
Author(s):  
Daqing Cui ◽  
Jeanett Low ◽  
Max Lundström ◽  
Kastriot Spahiu
Keyword(s):  
Cast Iron ◽  
Aqueous Phase ◽  
Atmospheric Oxygen ◽  
Near Field ◽  
Fission Products ◽  
Spent Fuel ◽  
Metal Foils ◽  
Oxygen Contamination ◽  
Leaching Experiment ◽  
Fuel Pin

ABSTRACTThe results of a spent fuel leaching experiment in which a fuel pin (17.7 g) was contacted with 380 mL of a 10 mM NaCl, 2 mM NaHCO3 solution by taking special care to minimize atmospheric oxygen contamination are presented. During the first 287 days, the fractions of inventory in the aqueous phase per day (f/d) increased nearly constantly for all nuclides (except for 100Mo), but were higher for fission products f/d(137Cs)=1.210−6, f/d(99Tc)=1.1·10−6 and f/d(90Sr)= 6.7 · 10−7 than for actinides: f/d (238U) =1.0 · 10−7, f/d(237Np)= 2.6 · 10−7 and f/d(239Pu) = 5.1 ·10−9. After adding iron, cast iron and copper foils (of ∼30 mm2 size), the concentrations of 238U, 237Np and 99Tc decreased by 80%, 97% and 88% to relatively stable levels (500ppb, 0.2 ppb and 0.6 ppb respectively). 239Pu concentrations increased from a level around 0.05 ppb to PuO2 solubility level, 0.5 ppb, and stabilized. The leaching process for 137Cs, 100Mo and 90Sr seems not to be influenced by the addition of metal foils. The observations in the present work contribute to an improved understanding of the behavior of spent fuel under near field repository conditions.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

Influence of Container Base Material (Fe) on SIMFUEL Leaching Behavior

2000 ◽  
Vol 663 ◽  
Author(s):  
J. Quiñones ◽  
J.A. Serrano ◽  
P.P. Díaz ◽  
J.L. Rodríguez Almazán ◽  
J. Cobos ◽  
...  
Keyword(s):  
Experimental Data ◽  
Redox Potential ◽  
Uranium Concentration ◽  
Near Field ◽  
Base Material ◽  
Fission Products ◽  
Sharp Decrease ◽  
Spent Fuel ◽  
Leaching Behavior ◽  
Redox Sensitivity

ABSTRACTThe chemical stability of spent fuel will be greatly influenced by the redox potential of the near field. Presence of reductants such as iron is likely to be an important factor to maintain the original integrity of spent fuel. In this work experimental data about the influence of metallic iron (container base material) on SIMFUEL leaching behavior under simulated granite and saline repository conditions is presented. In the presence of iron uranium concentration undergoes a sharp decrease. This is much more noticeable in the experiments performed under initial oxic conditions. The effect of iron on simulated fission products of SIMFUEL is very important for the elements with high redox sensitivity such us molybdenum. On the contrary, strontium remains stable during the entire tests and it seems not be affected by changes in redox potential.

Zirconia Inert Matrix for Plutonium Utilisation and Minor Actinides Disposition in Thermal Reactors

2006 ◽  
Vol 45 ◽  
pp. 1907-1914
Author(s):  
Claude Degueldre
Keyword(s):  
Solid Solution ◽  
Ion Irradiation ◽  
Near Field ◽  
Fission Products ◽  
Heavy Ion ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix

The toxicity of the UO2 spent fuel is dominated by plutonium and minor actinides (MA): Np, Am and Cm, after decay of the short live fission products. Zirconia ceramics containing Pu and MA in the form of an Inert Matrix Fuel (IMF) could be used to burn these actinides in Light Water Reactors. Optimisation of the fuel designs dictated by properties such as thermal, mechanical, chemical and physical must be performed with attention for their behaviour under irradiation. Zirconia must be stabilised by yttria to form a solid solution such as AnzYyPuxZr1-yO2-y where minor actinide oxides are also soluble. Burnable poison may be added if necessary such as Gd, Ho, Er, Eu or Np, Am them-self. These cubic solid solutions are stable under heavy ion irradiation. The retention of fission products in zirconia, under similar thermodynamic conditions, is a priori stronger, compared to UO2, the lattice parameter being larger for UO2 than for (Y,Zr)O2-x. (Er,Y,Pu,Zr)O2-x in which Pu contains 5% Am was successfully irradiated in the Proteus reactor at PSI, in the HFR facility, Petten as well as in the Halden Reactor. These irradiations make the Swiss scientists confident to irradiate such IMF in a commercial reactor that would allow later a commercial deployment of such a fuel for Pu and MA utilisation in a last cycle. The fuel forms namely pellet of solid solution, cercer or cermet fuel are discussed considering the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. As spent fuels these IMF’s are demanding materials from the solubility point of view, this parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is 106 times smaller than glass, which makes the zirconia material very attractive for deep geological disposal. The desired objective would be to use IMF to produce energy in reactors, opting for an economical and ecological solution.

Chemical Interactions in the Near-Field of a Spent Fuel Repository - Preliminary Results from a Long-Term Laboratory Experiment

2002 ◽  
Vol 713 ◽  
Author(s):  
H. Aalto ◽  
T. Carlsson ◽  
H. Kumpulainen ◽  
J. Lehikoinen ◽  
A. Muurinen
Keyword(s):  
Cast Iron ◽  
Laboratory Experiment ◽  
Near Field ◽  
External Solution ◽  
Corrosion Products ◽  
Spent Fuel ◽  
Iron Corrosion ◽  
N2 Atmosphere ◽  
External Water ◽  
Set Up

ABSTRACTA ten-year-long laboratory experiment was recently started in order to study interactions between the groundwater, the bentonite, and the canister in the near-field of the planned Finnish repository for spent nuclear waste. The experimental set-up consists of a number of samples containing sodium MX-80 bentonite and cast iron cylinders placed in copper vessels. The bentonite stays in contact with an external solution, either distilled water or a 0.5 M NaCl solution, via metal sinters. The experiments are performed under anoxic conditions (N2 atmosphere) at ambient room temperature (∼30 °C). The experimental measurements focus mainly on the development of (1) the water chemistry, (2) the cast iron corrosion, and (3) the diffusion of corrosion products inside the bentonite, since these processes are the most likely ones to produce detectable changes during the ten-year-long programme.The first analyses, after 9 months, showed that the iron corrosion had started and that the corrosion products had migrated into the bentonite. The corrosion products could not be identified by XRD, which either meant that the material was amorphous or present in concentrations too low to be detectable. The analyses of the gas phase indicated the presence of hydrogen, which most probably was the result of the iron corrosion. The compositions of the external water and the porewater were determined with regard to the major species of interest. A closer evaluation of the results will not be given until more data are at hand.

The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
Shirley A. Rawson ◽  
William L. Neal ◽  
James R. Burnell
Keyword(s):  
Low Carbon Steel ◽  
Fission Products ◽  
Nuclear Waste Repository ◽  
Low Carbon ◽  
Spent Fuel ◽  
Hydrothermal Conditions ◽  
Waste Package ◽  
Dissolved Species ◽  
Radionuclide Release ◽  
Time Invariant

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.

Spent Fuel: Characterization Studies and Dissolution Behaviourunder Disposal Conditions

1987 ◽  
Vol 112 ◽  
Author(s):  
L. H. Johnson ◽  
D. W. Shoesmith ◽  
S. Stroes-Gascoyne
Keyword(s):  
Release Kinetics ◽  
Fission Products ◽  
Solution Chemistry ◽  
Future Research ◽  
Spent Fuel ◽  
Research Areas ◽  
Dissolution Model ◽  
Underground Disposal ◽  
Radionuclide Release

AbstractThe concept of disposal of unreprocessed spent fuel has now been under study internationally for over ten years. Considerable progress has been made in understanding the factors that will control radionuclide release from spent fuel in an underground disposal vault. This progress is reviewed and the research areas of significance in providing further data for source term models are discussed. Key areas for future research are identified; these include improved characterization of spent fuel to determine the inventories of fission products at grain boundaries, together with their release kinetics; and a better understanding of the effects of solution chemistry on spent fuel dissolution, in particular the effects of salinity, redox chemistry, and radiolysis of groundwater. Approaches to modelling the dissolution of spent fuel are discussed, and a possible approach for developing an oxidative dissolution model is outlined.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

The geological research in France - The Dosssier 2005 Argile

2006 ◽  
Vol 985 ◽  
Author(s):  
Frederic Plas ◽  
Jacques WENDLING
Keyword(s):  
Near Field ◽  
Concept Design ◽  
Dose Calculations ◽  
Spent Fuel ◽  
Geological Medium ◽  
Source Data ◽  
Underground Research Laboratory ◽  
Broad Outline ◽  
Geological Research

AbstractAt the end of fifteen years of researchs defined by the French act of December 30, 1991 on radwaste management, Andra gave a report, “Dossier Argile 2005”, which concluded with the feasibility of a reversible disposal in the argillaceous Callovo-Oxfordien formation studied by means of an underground research laboratory at Meuse/Haute-Marne site. Starting from source data like the characteristics of the geological medium and the waste inventory, the process followed by Andra to achieve at this conclusion is of type sequential and iterative between concept design, scientific knowledge, in particular that of the phenomenological evolution of the reposiroty and its geological environment from operating period to long term, and Safety assessment. The “Dossier Argile 2005” covers a broad radwaste inventory, ILLW, HLW and Spent Fuel, so that it makes it possible to cover whole of the technological, scientific and safety topics. This article will give an overview of the geological disposal studies in France and draw the main conclusion of the Dossier 2005 Argile. It will be focused on the near field (Engineering components and near field host rock), while considering if necessary its integration within the whole system. After a short description of the concepts (incl. waste inventory and the characteristics of the Meuse/Haute the Marne site) and the functions of the components of repository and geological medium, one will describe successively the broad outline of the phenomenological evolution of repository and the geological medium in near field, by in particular releasing the time scales of processes and uncertainties of knowledge. On this basis, one will indicate the safety scenarios which were considered and the broad outline of performance and dose calculations. Lessons learn from the Dossier 2005 Argile will be discussed and perspective and priority for future will be indicated.

Nondestructive, Energy-Dispersive, X-Ray Fluorescence an Analysis of Actinide Stream Concentrations from Reprocessed Nuclear Fuel

1979 ◽  
Vol 23 ◽  
pp. 163-176
Author(s):  
D. C. Camp ◽  
W. D. Ruhter
Keyword(s):  
Nuclear Fuel ◽  
Nitrate Solution ◽  
Fission Products ◽  
Analytical Techniques ◽  
Fluorescence Analysis ◽  
Energy Dispersive ◽  
Light Water Reactors ◽  
Spent Fuel ◽  
X Ray ◽  
Water Reactors

In the event that nuclear fuel from light water reactors (LWR) is reprocessed to reclaim the uranium or plutonium, several analytical techniques will be used for product accountability. Generally, the isotopic content of both the plutonium and uranium in the reprocessed product will have to be accurately determined. One plan for the reprocessing of LWR spent fuel incorporates the following scheme. After separation from both the fission products and transplutonium actinides (including neptunium and americium), part of the uranium and all of the plutonium in a nitrate solution will merge together to form a coprocessed stream. This solution will be concentrated by evaporation and sent to a hold tank for accountability. Input concentrations into the hold tank could be up to 350 g U/ℓ and nearly 50 g Pu/ℓ. The variation to be expected in these concentrations is not known. The remaining uranium fraction will be further purified and sent to a separate storage tank. Its expected stream concentration will be about 60 g U/ℓ. These two relatively high actinide stream concentrations can be monitored rapidly, quantitatively, and nondestructively using the technique of energy-dispersive x-ray fluorescence analysis(XRFA).

Sign in / Sign up

Export Citation Format

Share Document