Preparation and characterization of UO2-based AGR SIMFuel

2014 ◽  
Vol 1665 ◽  
pp. 245-251 ◽  
Author(s):  
Zoltan Hiezl ◽  
David Hambley ◽  
William E. Lee
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Product Distribution ◽  
Sintering Process ◽  
Secondary Phases ◽  
Fuel Pellets ◽  
Stable Elements ◽  
Grey Phase ◽  
Good Agreement

ABSTRACTPreparation and characterization of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the Fispin code provided by NNL. SIMFuel pellets exhibit a microstructure up to 92% TD. During the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE)O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF, although they are smaller in size than those examined in this study. The grain size of the produced SIMFuel is in good agreement with literature references.

scholarly journals An Automated SeaFAST ICP-DRC-MS Method for the Determination of 90Sr in Spent Nuclear Fuel Leachates

Molecules ◽  
2020 ◽  
Vol 25 (6) ◽  
pp. 1429 ◽  
Author(s):  
Víctor Vicente Vilas ◽  
Sylvain Millet ◽  
Miguel Sandow ◽  
Luis Iglesias Pérez ◽  
Daniel Serrano-Purroy ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Analytical Approach ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Chemical Purification ◽  
Radiochemical Method ◽  
Good Agreement

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

scholarly journals XANES characterization of UO2/Mo(Pd) thin films as models for ϵ-particles in spent nuclear fuel

2013 ◽  
Vol 430 ◽  
pp. 012113 ◽  
Author(s):  
M A Denecke ◽  
T Petersmann ◽  
R Marsac ◽  
K Dardenne ◽  
T Vitova ◽  
...  
Keyword(s):  
Thin Films ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel

Studying the Progress of Sintering in Ferrous Powder Compacts by In-Situ Measuring the Thermal Conductivity

2018 ◽  
Vol 18 (2) ◽  
pp. 80-95
Author(s):  
H. Danninger ◽  
G. Leitner ◽  
Ch. Gierl-Mayer
Keyword(s):  
Thermal Conductivity ◽  
Laser Flash ◽  
Sintering Process ◽  
In Situ Characterization ◽  
Dimensional Changes ◽  
Flow Through ◽  
Powder Compacts ◽  
Good Agreement

Abstract In situ characterization of the sintering process is a difficult task, in particular for systems without pronounced dimensional changes. Dilatometry is not too helpful in those cases, and therefore other properties have to be recorded. In the present study, sintering of ferrous powder compacts was studied in situ by measuring the thermal diffusivity a using a laser flash apparatus. This property is a measure to characterise the heat flow through a material; it depends on the contact area between the particles and thus reveals their change during sintering. It is shown that the change of a during sintering of ferrous compacts is much less pronounced than in the case of cemented carbides which is not surprising when regarding the widely differing porosity changes. The results are however in good agreement with expectations when considering some experimental limitations. The trend for the thermal conductivity λ. which can be calculated from a, the specific heat and the density, is in good agreement with that found for the electrical conductivity, both properties being linked through Wiedemann-Franz’ law.

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.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

Characterization of alloy particles extracted from spent nuclear fuel

2012 ◽  
Vol 420 (1-3) ◽  
pp. 328-333 ◽  
Author(s):  
D. Cui ◽  
V.V. Rondinella ◽  
J.A. Fortner ◽  
A.J. Kropf ◽  
L. Eriksson ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Alloy Particles
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