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.

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.

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

Behaviour of CS and I In Zirconia Based Fuel

1998 ◽  
Vol 540 ◽  
Author(s):  
C. Degueldre ◽  
M. Pouchon ◽  
M. Doebli ◽  
G. Ledergerber
Keyword(s):  
Incident Energy ◽  
Nuclear Reactors ◽  
Fission Products ◽  
Concentration Profiles ◽  
Reactor Safety ◽  
Inert Matrix Fuel ◽  
Light Water ◽  
Inert Matrix ◽  
Stepwise Heating ◽  
Implantation Depth

AbstractA zirconia based ceramic is foreseen as an inert matrix fuel for burning excess plutonium in light water nuclear reactors. For reactor safety reasons the behaviour of volatile fission products such as cesium and iodine must be studied since a retention of fission products is favourable for licensing the studied inert matrix fuel. In this study, implantation of Cs and I was performed into polycrystalline (Zr0.85, Y0.15)O1.925 samples. The implantation depth was selected on the basis of the ability to observe by Rutherford backscattering spectroscopy (RBS) the behaviour of Cs and I after treatment. With a 1 MeV incident energy, the ions are implanted at a depth of 200 nm as predicted by TRIM. After implantations full quantification of I and Cs concentration profiles was performed by RBS. The implantation profiles are measured as a function of sample temperature during stepwise heating programs. It is interesting to observe retention of Cs and I at relatively high temperature (e.g. for 2 h, below 900 K for Cs and below 1400 K for I). This behaviour is likely to be due to the size and interactions of these species in the zirconia solid solution.

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.

Leaching of Spent Fuel under Anaerobic and Reducing Conditions

2002 ◽  
Vol 757 ◽  
Author(s):  
Yngve Albinsson ◽  
Arvid Ödegaard-Jensen ◽  
Virginia M. Oversby ◽  
Lars O. Werme
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Fluid Phase ◽  
Inert Atmosphere ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Reducing Conditions ◽  
Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

The Influence of Ligands and Precipitates on the Release of Nuclides from the Near Field under Natural Repository Conditions

2000 ◽  
Vol 663 ◽  
Author(s):  
L. Liu ◽  
I. Neretnieks
Keyword(s):  
Uranium Oxide ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Radiation Power ◽  
Near Field ◽  
Reference Scenario ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Fuel Surface ◽  
Fuel Pellets

ABSTRACTOnce groundwater intrudes into a damaged canister and wets the spent fuel pellets, radiation emitted from the spent nuclear fuel splits nearby water into oxidizing and reducing species. This may lead to an oxidizing condition near the fuel pellets. As a result, uranium oxide that makes up the fuel matrix will become more soluble, and the incorporated radionuclides will be released more rapidly. The dissolution process is, however, a dynamic one that can be influenced by many factors. Of great importance are the radiation power of the fuel matrix, the concentration of ligands near the fuel surface, and the transport resistance of the near field. Consequently, the escape of nuclides from the damaged canister is dominated mainly by the intrusion of ligands, and the precipitation/dissolution of secondary phases within the fuel rods. To investigate the possible effects of ligands and precipitates, a coupled dissolution and transport model, which includes the barrier effect of the Zircaloy claddings, is developed. The application of the model to a SKB-specified reference scenario indicates that by far the largest fraction of the oxidized uranium will reprecipitate within the canister. This may significantly decrease the fuel surface available for oxidation and the water available for radiolysis. Subsequently, much less fuel matrix will be dissolved and much less of the other nuclides will be released. Simulations further identify that carbonate and silicate have the greatest influences on the formation of secondary phases, and on the release of nuclides, under natural repository conditions.

scholarly journals Partitioning and transmutation contribution of MYRRHA to an EU strategy for HLW management and main achievements of MYRRHA related FP7 and H2020 projects: MYRTE, MARISA, MAXSIMA, SEARCH, MAX, FREYA, ARCAS

2020 ◽  
Vol 6 ◽  
pp. 33
Author(s):  
Hamid Aït Abderrahim ◽  
Peter Baeten ◽  
Alain Sneyers ◽  
Marc Schyns ◽  
Paul Schuurmans ◽  
...  
Keyword(s):  
Waste Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Economic Feasibility ◽  
Potential Contribution ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Deep Geological Disposal ◽  
Partitioning And Transmutation

Today, nuclear power produces 11% of the world's electricity. Nuclear power plants produce virtually no greenhouse gases or air pollutants during their operation. Emissions over their entire life cycle are very low. Nuclear energy's potential is essential to achieving a deeply decarbonized energy future in many regions of the world as of today and for decades to come, the main value of nuclear energy lies in its potential contribution to decarbonizing the power sector. Nuclear energy's future role, however, is highly uncertain for several reasons: chiefly, escalating costs and, the persistence of historical challenges such as spent fuel and radioactive waste management. Advanced nuclear fuel recycling technologies can enable full use of natural energy resources while minimizing proliferation concerns as well as the volume and longevity of nuclear waste. Partitioning and Transmutation (P&T) has been pointed out in numerous studies as the strategy that can relax constraints on geological disposal, e.g. by reducing the waste radiotoxicity and the footprint of the underground facility. Therefore, a special effort has been made to investigate the potential role of P&T and the related options for waste management all along the fuel cycle. Transmutation based on critical or sub-critical fast spectrum transmuters should be evaluated in order to assess its technical and economic feasibility and capacity, which could ease deep geological disposal implementation.

Modeling spent nuclear fuel alteration and radionuclide migration in disposal conditions

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Laurent de Windt ◽  
H. Schneider ◽  
C. Ferry ◽  
H. Catalette ◽  
V. Lagneau ◽  
...  
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Near Field ◽  
Radioactive Decay ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Physico Chemical ◽  
Integrated Performance

A physico-chemical model developed for spent fuel alteration was integrated in a global reactive transport model of a spent fuel disposal system, considering both decaying and stable isotopes, corroded steel canisters, bentonite backfills and a clayey host-rock. Fuel evolution took into account radiolytic-enhanced corrosion and long-term solubility-controlled dissolution as well as instantaneous release fractions. The calculations show that spent-fuel dissolution has no significant alteration effect on the near-field components except an oxidizing plume in the vicinity of the waste packages. The dissolved uranyl species, partly precipitate as schoepite on the fuel pellets, and partly diffuse in the near-field where magnetite and pyrite reduce U(VI) to yield uraninite precipitation. Under disposal conditions, preliminary calculations indicate that steel corrosion may generate sufficient dissolved hydrogen as to react with radiolytic oxidants and inhibit fuel dissolution. The formation of a protective schoepite layer could also reduce the alteration of fuel pellets. Radionuclides migration (Am, Cs, I) in the near-field is discussed in a second stage discriminating between sorption, precipitation and radioactive decay processes. The migration of Cs is translated in terms of cumulative activity profiles useful for integrated performance assessment.

Study of SIMFUEL corrosion under hyper-alkaline conditions in the presence of silicate and calcium

MRS Advances ◽  
2016 ◽  
Vol 2 (10) ◽  
pp. 543-548
Author(s):  
Alexandra Espriu-Gascon ◽  
David W. Shoesmith ◽  
Javier Giménez ◽  
Ignasi Casas ◽  
Joan de Pablo
Keyword(s):  
Corrosion Potential ◽  
Oxidation Process ◽  
Spent Fuel ◽  
Cyclic Voltammetric ◽  
Geological Disposal ◽  
Similar Composition ◽  
X Ray ◽  
Deep Geological Disposal ◽  
Alkaline Conditions ◽  
Main Components

ABSTRACTCement has been considered as a possible material present in the Deep Geological Disposal (DGD) [1] . In order to determine the effect of cementitious waters on the oxidation of the surface of Spent Fuel (SF), a series of electrochemical experiments were performed, to study the influence of two main components of cementitious water: calcium and silicate.Test solutions with Na2SiO3 and/or CaCl2 were prepared at pH 12 and NaCl 0.1 mol·dm-3 as ionic medium. A 3 at.% doped SIMFUEL was used to perform cyclic voltammetric (CV), potentiostatic and corrosion potential (ECORR) experiments. After potentiostatic and ECORR experiments, the SIMFUEL surface was analyzed using X-Ray Photoelecton Spectroscopy (XPS).The results showed that the presence of silicate decreased the SIMFUEL oxidation between -100 mV and 300 mV. When Ca2+ was added, the whole oxidation process was shifted to higher potentials which indicated a protective effect of the combination of Ca2+ and SiO32- . The XPS results obtained after potentiostatic experiments at 200 mV showed that the presence of silicate partially suppressed the oxidation of SIMFUEL, as indicated by the contribution of both U(IV) and U(V) XPS to the U 4f7/2 band (∼ 38%). After the addition of calcium, the predominant uranium oxidized state contribution on the surface was U(V) (40%). After the ECORR experiments, the ECORR values were similar either with or without silicate in solution (-80 mV and -70 mV respectively). The resulting surface also exhibited a similar composition. When calcium was added to the electrolyte, the ECORR value was suppressed to -105 mV, and XPS showed that the surface was less oxidized than with the other two electrolytes.

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