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 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.

The Influence of Near Field Redox Conditions on Spent Fuel Leaching

2002 ◽  
Vol 713 ◽  
Author(s):  
Kastriot Spahiu ◽  
Ulla-Britt Eklund ◽  
Daqing Cui ◽  
Max Lundström
Keyword(s):  
Hydrogen Pressure ◽  
Near Field ◽  
Free Water ◽  
Spent Fuel ◽  
Leaching Behavior ◽  
Simulated Groundwater ◽  
Dissolved Hydrogen ◽  
Fuel Components ◽  
Very High ◽  
Anaerobic Environment

ABSTRACTIn a repository, the spent fuel could come in contact with groundwater if the canister or container has breached. The system may be quite complex with oxygen-free water, uranium dioxide, a corroding metal, such as iron, and a radiation field present at the same time. In an anaerobic environment iron and mild steel will corrode and hydrogen will be evolved. The equilibrium hydrogen pressure for this reaction is very high. At some time after water intrusion, there will be large amounts of dissolved hydrogen in the near field, corresponding to a partial pressure at least equivalent to the hydrostatic pressure at the repository depth. For this reason, we investigated the leaching behavior of 0.25-0.5 mm sized fragments of PWR spent fuel (43 MWd / Kg U) in simulated groundwater solution (10 mM NaCl and 2 mM HCO3-) under 5 MPa hydrogen and argon pressure. In a leaching experiment under 5 MPa hydrogen at 25 °C, the total U concentration was found to be <10−8 M. After refilling of the autoclave with new solution at 70°C, the total U concentration first increased to 10−6.3M, and then quickly decreased to 10−8 M. The leaching behavior of uranium and other fuel components indicates that under pressurized hydrogen, the spent fuel dissolution is substantially hindered. Leaching results obtained after the substitution of hydrogen by argon at the same pressure and temperature are also presented. Finally, some results on spent fuel leaching under pressurized argon are presented and comparatively discussed.

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.

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.

scholarly journals Revisiting the Analysis of the Isochronous Mass Measurements of Uranium Fission Fragments at the ESR

2020 ◽  
Vol 227 ◽  
pp. 02012
Author(s):  
R. S. Sidhu ◽  
R. J. Chen ◽  
Yu. A Litvinov ◽  
Y. H. Zhang ◽  
Keyword(s):  
Experimental Data ◽  
Data Analysis ◽  
State Of The Art ◽  
Fission Products ◽  
Mass Measurements ◽  
Fission Fragments ◽  
Uranium Fission

The re-analysis of experimental data on mass measurements of ura- nium fission products obtained at the ESR in 2002 is discussed. State-of-the-art data analysis procedures developed for such measurements are employed.

scholarly journals Rational design of new NO and redox sensitivity into connexin26 hemichannels

Open Biology ◽  
2015 ◽  
Vol 5 (2) ◽  
pp. 140208 ◽  
Author(s):  
Louise Meigh ◽  
Daniel Cook ◽  
Jie Zhang ◽  
Nicholas Dale
Keyword(s):  
Redox Potential ◽  
Rational Design ◽  
Salt Bridge ◽  
Disulfide Bridge ◽  
Wild Type ◽  
No Donors ◽  
Bridge Formation ◽  
Redox Sensitivity ◽  
Intracellular Redox

CO 2 directly opens hemichannels of connexin26 (Cx26) by carbamylating K125, thereby allowing salt bridge formation with R104 of the neighbouring subunit in the connexin hexamer. The formation of the inter-subunit carbamate bridges within the hexameric hemichannel traps it in the open state. Here, we use insights derived from this model to test whether the range of agonists capable of opening Cx26 can be extended by promoting the formation of analogous inter-subunit bridges via different mechanisms. The mutation K125C gives potential for nitrosylation on Cys125 and formation of an SNO bridge to R104 of the neighbouring subunit. Unlike wild-type Cx26 hemichannels, which are insensitive to NO and NO 2 − , hemichannels comprising Cx26 K125C can be opened by NO 2 − and NO donors. However, NO 2 − was unable to modulate the doubly mutated (K125C, R104A) hemichannels, indicating that an inter-subunit bridge between C125 and R104 is required for the opening action of NO 2 − . In a further test, we introduced two mutations into Cx26, K125C and R104C, to allow disulfide bridge formation across the inter-subunit boundary. These doubly mutated hemichannels open in response to changes in intracellular redox potential.

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.

Aqueous leaching of ADOPT and standard UO2 spent nuclear fuel under H2 atmosphere

MRS Advances ◽  
2020 ◽  
Vol 5 (3-4) ◽  
pp. 167-175
Author(s):  
Alexandre Barreiro Fidalgo ◽  
Olivia Roth ◽  
Anders Puranen ◽  
Lena Z. Evins ◽  
Kastriot Spahiu
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Preliminary Data ◽  
Spent Nuclear Fuel ◽  
Uranium Concentration ◽  
Hydrogen Atmosphere ◽  
Dissolution Behavior ◽  
Leaching Behavior ◽  
New Type ◽  
Failure Scenario

ABSTRACTLeaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

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.

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