Assessing the role of spent fuel surfaces during leaching in presence of hydrogen by using Cr(VI) as a redox marker

2012 ◽  
Vol 1475 ◽  
Author(s):  
A. Puranen ◽  
E. Ekeroth ◽  
M. Granfors ◽  
J. Low ◽  
K. Spahiu
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Spent Fuel ◽  
Low Concentrations ◽  
Metallic Inclusions ◽  
Simulated Groundwater ◽  
Dark Green ◽  
Corrosion Of Iron ◽  
Sensitive Marker

ABSTRACTIn many deep repository concepts spent nuclear fuel (SNF) will be disposed in canisters containing large amounts of iron. Intrusion of groundwater in a failed canister may occur under the presence of hydrogen, expected to be produced by the anoxic corrosion of iron. Compelling evidence now exists that hydrogen inhibits oxidative dissolution of SNF, the mechanism is however not fully understood. Hydrogen generally requires a catalyst in order to operate as a reductant. The metallic inclusions (ε-particles) present in SNF are a likely catalyst for this process due to their noble metal content. There is also evidence that the SNF UO2 matrix or doping of the UO2 with fission products can activate hydrogen. In most spent fuel experiments carried out under hydrogen, a decrease in concentration of all redox sensitive nuclides originating from a pre-oxidized layer is observed. Given their low concentrations and abundance in the fuel, it has however been difficult to detect any reductive precipitation on the fuel surfaces.In this study, Cr(VI) oxyanions were employed as a redox sensitive marker, as Cr(VI) is expected to precipitate as Cr(III) oxide on the catalyst that activates hydrogen.In the experiments PWR spent fuel (43 MWd/kgU) was leached in simulated groundwater (10 mM NaCl, 2 mM NaHCO3) at 25 and 70 ◦C under 5 MPa of hydrogen and dissolved Cr(VI). Dark green, Cr(III)-oxide was found to precipitate; mapping by electron microscopy (SEM-WDS) evidenced a Cr rich layer covering the fuel, suggesting that the whole fuel surface is catalyzing the reduction of chromium.

Safety Assessment of Fuel Storage Ponds Following Cooling Faults

2016 ◽  
Author(s):  
Jonathan Webb ◽  
Charles Bridgford
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Nuclear Safety ◽  
Nuclear Industry ◽  
Spent Fuel ◽  
Cooling Systems ◽  
External Cooling ◽  
Safety Case

For spent nuclear fuel stored within a cooling pond, the essential nuclear safety functions of control, cooling and containment are fulfilled by maintaining an appropriate depth of water above the fuel. External cooling systems remove the decay heat generated by the spent fuel stored within the pond, in order to maintain the temperature of the water at a constant level. In the event of a fault within these external cooling systems, there is the potential for a temperature excursion within the pond. Historically the UK nuclear industry has considered that such faults would pose no threat to the structural integrity of the pond containment and hence the only loss of water would be due to evaporation following a loss of cooling. However, more recently, it has been recognised that such temperature excursions may result in through-wall cracking leading to a loss of water and undermining of these essential safety functions. This paper outlines the safety case implications of these realisations and the way in which they are being addressed within the UK’s nuclear power stations. The paper considers the effects of thermal transient faults on the concrete pond structure and the potential nuclear safety issues which may occur as a result of this. In response to potential pond cooling faults, consideration is given to the requirement for engineered protection systems along with the safety role of the operator in identifying and responding to faults of this kind. Operators provide a versatile mechanism for identifying fault conditions and taking remedial actions, however, the benefit which can be formally claimed for their role within a safety case is generally limited by the availability or reliability of instrumentation to reveal a fault condition. Post fault operator actions may also be limited by the timescales available following a fault, or by other demands on the operators, which may occur in the event of an external hazard which affects multiple site systems. To quantify the timescales available for post fault remedial action, it is necessary to quantify the rate of water loss from the pond, along with the relationship between pond water depth and the radiological consequences both on-site and off-site. This paper investigates the difficulties which may be encountered in quantifying the role of post fault operator actions within such a safety case, and in demonstrating that the overall nuclear safety risk is acceptably low and as low as reasonably practicable (ALARP).

Modeling of UO2 Aqueous Dissolution Over a Wide Range of Conditions

1993 ◽  
Vol 333 ◽  
Author(s):  
Steven A. Steward ◽  
Homer C. Weed
Keyword(s):  
Activation Energy ◽  
Dissolved Oxygen ◽  
Fission Products ◽  
Statistical Experimental Design ◽  
Power Dependence ◽  
Spent Fuel ◽  
Wide Range ◽  
Simulated Groundwater ◽  
Effects Of Temperature ◽  
Half Power

ABSTRACTPreviously it was not possible to predict reliably the rate at which spent fuel would react with groundwater because of conflicting data in the literature. The dissolution of the UO2 spent fuel matrix is a necessary step for aqueous release of radioactive fission products. Statistical experimental design was used to plan a set of UO2 dissolution experiments to examine systematically the effects of temperature (25-75°C), dissolved oxygen (0.002-0.2 atm overpressure), pH (8-10) and carbonate (2-200x10-4 molar) concentrations on UO2 dissolution. The average dissolution rate was 4.3 mg/m2/day. The regression fit of the data indicate an Arrhenius type activation energy of -8750 cal/mol·K and a half-power dependence on dissolved oxygen in the simulated groundwater.

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.

Distribution of Palladium During Spent Fuel Reprocessing

2003 ◽  
Author(s):  
Yu. Pokhitonov ◽  
V. Romanovski ◽  
P. Rance
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Platinum Group Metals ◽  
Spent Fuel ◽  
Platinum Metals ◽  
Spent Fuel Reprocessing ◽  
Platinum Group ◽  
Fuel Reprocessing

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.

scholarly journals The Limitations of an Air-Oxidation Breakaway Model to Predict a Zirconium Fire in a Spent Nuclear Fuel Pool Accident

2019 ◽  
Vol 11 (22) ◽  
pp. 6364
Author(s):  
Sanggil Park ◽  
Min Bum Park
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Phenomenon ◽  
Fission Products ◽  
Model Parameters ◽  
Spent Fuel ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Spent Fuel Pool

The OECD/NEA Spent Fuel Pool (SFP) project was conducted to investigate consequences of spent nuclear fuel pool accident scenarios. From the project, it was observed that cladding temperature could abruptly increase at a certain point and the cladding was completely oxidized. This phenomenon was called a “zirconium fire”. This zirconium fire is one of the crucial concerns for spent fuel pool safety under a postulated loss of coolant accident scenario, since it would lead to an uncontrolled mass release of fission products into the environment. To capture this critical phenomenon, an air-oxidation breakaway model has been implemented in the MELCOR code. This study examines this air-oxidation breakaway model by comparing the SFP project test data with a series of MELCOR code sensitivity calculation results. The air-oxidation model parameters are slightly altered to investigate their sensitivities on the occurrence of the zirconium fire. Through such sensitivity analysis, limitations of the air-oxidation breakaway model are identified, and needs for model improvement is recommended.

scholarly journals Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor

2019 ◽  
Vol 5 (4) ◽  
pp. 337-343
Author(s):  
Sergey N. Ivanov ◽  
Sergey I. Porollo ◽  
Yury D. Baranaev ◽  
Vladimir F. Timofeev ◽  
Yury V. Kharizomenov
Keyword(s):  
Dose Rate ◽  
Water Samples ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Fuel Composition ◽  
Spent Fuel ◽  
Artificial Defect ◽  
Corrosion Tests ◽  
Long Term Storage ◽  
Fuel Elements

Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.

The Role of the Actinides in a Performance Assessment of a Nuclear Waste Repository. SKB's Supporting Actinide Research

2010 ◽  
Vol 1265 ◽  
Author(s):  
Lars Werme ◽  
Sergei Butorin ◽  
Peter M Oppeneer
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Performance Assessments ◽  
Nuclear Waste Repository ◽  
Geologic Repository ◽  
Activation Products ◽  
Waste Repository

AbstractAfter a few hundred years, the actinides will dominate the radiotoxicity of spent nuclear fuel. This does not necessarily mean that the actinides will dominate the dose to organisms at the surface above a geologic repository. Quite the contrary, in most performance assessments this dose is dominated by long-lived fission products, activation products and, in the very long perspective, actinide daughters.This makes the far-field migration properties of the actinides less interesting for further research. There are, however, other aspects of the presence of actinides in spent nuclear fuel and some of these and SKB's research in these fields is presented and discussed here.

Synthesis of Uranium Phosphate Phases and Potential Retardation Effects on Spent Fuel Radionuclide Migration

2002 ◽  
Vol 757 ◽  
Author(s):  
A. S. Turner ◽  
D. J. Wronkiewicz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Potassium Phosphate ◽  
Fission Products ◽  
Chemical Compositions ◽  
Spent Fuel ◽  
Variable Chemical ◽  
Calcium Phosphorus ◽  
Uranyl Phosphates ◽  
Uranium Phosphate

ABSTRACTThe UO2 in spent nuclear fuel is unstable in the oxidizing conditions within the volcanic tuffs at the proposed nuclear repository at Yucca Mountain, Nevada. Over time, the UO2 will oxidize and corrode, releasing actinides and fission products to the surrounding environment. However, uranyl (U6+) phosphates (autunite, phurcalite, sodium autunite, etc.) are stable in such an oxidizing environment. The mobility of released radionuclides may be greatly retarded if they can be incorporated into these naturally stable phosphate phases, while the complex structures, variable chemical compositions, and natural analogue occurrences of the uranyl phosphates suggests such a process is favorable. Current tests have focused on synthesizing such phases by reacting uranium oxynitrate or UO3 with a calcium, sodium, or potassium phosphate and a base (if necessary) in a Teflon reaction vessel. Excess water is added, and the solution is heated at 90°C for 7, 35, or 182 days. SEM analyses have confirmed that various uranium phosphate crystalline solids have formed. XRD results indicate that tests using two different calcium phosphorus source materials, Ca2P2O7 and Ca10(OH)2(PO4)6, have both created synthetic phosphuranylite, Ca(UO2)[(UO2)3(OH)2(PO4)2] 2*12H2O. The formation of this phase is appears to be kinetically favored over other similar phases. Results utilizing sodium phosphates, NaH2(PO4)*H2O and Na2HPO4, have produced sodium autunite (NaUO2PO4*2H2O), but other phases are probably also present. Test results utilizing potassium phosphate, K3PO4, were inconclusive. Experiments using surrogate radionuclides are currently being performed in order to determine whether radionuclides, such as 239Pu, 137Cs, and 99Tc, released from corroded spent nuclear fuel can become incorporated into the crystalline structure of specific uranium phosphate phases, effectively limiting any further migration.

scholarly journals Hot Isostatic Pressing of Ceramic Waste From Spent Nuclear Fuel

2002 ◽  
Author(s):  
Kenneth J. Bateman ◽  
Richard H. Rigg ◽  
James D. Wiest
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Hot Isostatic Pressing ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Spent Fuel ◽  
Isostatic Pressing ◽  
Ceramic Waste ◽  
Pressing Operation ◽  
Waste Salt

Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process.

Modelling of RBMK Spent Nuclear Fuel Characteristics

2014 ◽  
Author(s):  
Arturas Smaizys ◽  
Povilas Poskas ◽  
Ernestas Narkunas
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Experimental Investigations ◽  
Spent Fuel ◽  
Radionuclide Content ◽  
Fuel Assemblies ◽  
Fuel Characteristics ◽  
Final Shutdown ◽  
Rbmk Fuel

After the final shutdown of Ignalina NPP, total amount of spent nuclear fuel is approximately 22 thousands of fuel assemblies. Radionuclide content and its characteristics in spent fuel are initial data for analysis of various safety related areas such as shielding, thermal analysis, radioactive releases and other processes. Experimental investigations of radionuclide content and characteristics in spent nuclear fuel are complicated and expensive, therefore numerical evaluation methods are widely used. Numerical modelling of spent RBMK fuel characteristics was performed using TRITON code from SCALE 6.1 system. Activities of fission products and actinides, gamma and neutron sources, decay heat obtained with TRITON code are compared with previous modelling results obtained using SAS2H sequence from the former SCALE 4.3 version. Some evaluated parameters are compared with published experimental data for RBMK spent nuclear fuel.

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