SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate.

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Leaching behaviour of UO2 containing α-emitting actinides

Radiochimica Acta ◽

10.1524/ract.2000.88.9-11.527 ◽

2000 ◽

Vol 88 (9-11) ◽

Cited By ~ 38

Author(s):

V.V. Rondinella ◽

Hj. Matzke ◽

Carlos J. Cobos ◽

T. Wiss

Keyword(s):

Radiation Field ◽

Nuclear Fuel ◽

Uranium Dioxide ◽

Spent Nuclear Fuel ◽

Spent Fuel ◽

Geologic Repository ◽

Fuel Surface ◽

Dissolution Behaviour ◽

Leaching Behaviour ◽

Irradiated Fuel

After a few hundred years in a geologic repository, α-emissions will dominate the radiation field in and around spent nuclear fuel. In the event of a failure of the spent fuel container, the dissolution of the uranium dioxide matrix in contact with groundwater could be enhanced by α-radiolysis, which could create oxidizing conditions at the fuel surface. The radioactivity (hence the radiolysis) of the irradiated fuel available nowadays is characterized by strong β- and γ-contributions, which are not representative of aged fuel in a repository. A possible way to single out the effects of α-radiolysis on the dissolution behaviour of irradiated fuel is to study unirradiated UO

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Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples

Volume 9: Student Paper Competition ◽

10.1115/icone25-67252 ◽

2017 ◽

Author(s):

Concettina Andrello ◽

Daniel Freis ◽

Rosa Lo Frano ◽

Dimitri Papaioannou ◽

Fabienne Delage

Keyword(s):

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Heat Rate ◽

Current Status ◽

High Level Waste ◽

Minor Actinide ◽

Spent Fuel ◽

Post Irradiation ◽

Irradiation Experiment

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation. MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option. In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds. Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination. After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.

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Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions

Nuclear Science and Technology ◽

10.53747/jnst.v9i4.134 ◽

2021 ◽

Vol 9 (4) ◽

pp. 16-26

Author(s):

Vinh Thanh Tran ◽

Thanh Mai Vu ◽

Van Khanh Hoang ◽

Viet Ha Pham Nhu

Keyword(s):

Comparative Analysis ◽

Fuel Assembly ◽

Spent Nuclear Fuel ◽

Minor Actinide ◽

Minor Actinides ◽

Spent Fuel ◽

Burnable Poison ◽

Uranium Fuel ◽

Enriched Uranium ◽

Actinide Transmutation

The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The comparative analysis was conducted for the two models of minor actinide loading in the LEU fuel assembly: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The parameters to be analysed and compared include the reactivity of the LEU fuel assembly versus burnup and the transmutation of minor actinide nuclides when loading different minor actinide vectors into the LEU fuel assembly.

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Reduction of Specimen Contamination in Electron-Beam Instruments

Proceedings, annual meeting, Electron Microscopy Society of America ◽

10.1017/s0424820100092840 ◽

1976 ◽

Vol 34 ◽

pp. 576-577

Author(s):

G. Lehmpfuhl ◽

P. J. Smith

Keyword(s):

Electron Beam ◽

Cold Trap ◽

Beam Diameter ◽

Transmission Electron ◽

Scanning Transmission ◽

Hydrocarbon Molecules ◽

Electron Microscopes ◽

Combination Of Methods ◽

Convergent Beam ◽

Very High

Specimens being observed with electron-beam instruments are subject to contamination, which is due to polymerization of hydrocarbon molecules by the beam. This effect becomes more important as the size of the beam is reduced. In convergent-beam studies with a beam diameter of 100 Å, contamination was observed to grow on samples at very high rates. Within a few seconds needles began forming under the beam on both the top and the underside of the sample, at growth rates of 400-500 Å/s, severely limiting the time available for observation. Such contamination could cause serious difficulty in examining a sample with the new scanning transmission electron microscopes, in which the beam is focused to a few angstroms.We have been able to reduce the rate of contamination buildup by a combination of methods: placing an anticontamination cold trap in the sample region, preheating the sample before observation, and irradiating the sample with a large beam before observing it with a small beam.

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In-Situ TEM Annealing Observation of Helium Bubble Evolution in Pre-Irradiated FeCoNiCrTi0.2 Alloys

Materials ◽

10.3390/ma14133727 ◽

2021 ◽

Vol 14 (13) ◽

pp. 3727

Author(s):

Huanhuan He ◽

Zhiwei Lin ◽

Shengming Jiang ◽

Xiaotian Hu ◽

Jian Zhang ◽

...

Keyword(s):

Vacuum Arc ◽

In Situ Tem ◽

Face Centered Cubic ◽

Helium Bubble ◽

High Entropy ◽

Helium Bubbles ◽

Transmission Electron ◽

Bubble Evolution ◽

Face Centered

The FeCoNiCrTi0.2 high-entropy alloys fabricated by vacuum arc melting method, and the annealed pristine material, are face centered cubic structures with coherent γ’ precipitation. Samples were irradiated with 50 keV He+ ions to a fluence of 2 × 1016 ions/cm2 at 723 K, and an in situ annealing experiment was carried out to monitor the evolution of helium bubbles during heating to 823 and 923 K. The pristine structure of FeCoNiCrTi0.2 samples and the evolution of helium bubbles during in situ annealing were both characterized by transmission electron microscopy. The annealing temperature and annealing time affect the process of helium bubbles evolution and formation. Meanwhile, the grain boundaries act as sinks to accumulate helium bubbles. However, the precipitation phase seems have few effects on the helium bubble evolution, which may be due to the coherent interface and same structure of γ’ precipitation and matrix.

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Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

MRS Proceedings ◽

10.1557/proc-985-0985-nn01-03 ◽

2006 ◽

Vol 985 ◽

Author(s):

Jeffrey A. Fortner ◽

A. Jeremy Kropf ◽

James L. Jerden ◽

James C. Cunnane

Keyword(s):

Transition Zone ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Aqueous Environment ◽

Spent Fuel ◽

Aqueous Corrosion ◽

Reduction Potentials ◽

Chemical Effects ◽

Before And After ◽

X Ray Absorption

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

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The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽

10.1557/adv.2018.231 ◽

2018 ◽

Vol 3 (19) ◽

pp. 991-1003 ◽

Cited By ~ 3

Author(s):

Evaristo J. Bonano ◽

Elena A. Kalinina ◽

Peter N. Swift

Keyword(s):

United States ◽

Nuclear Fuel ◽

United States Of America ◽

Spent Nuclear Fuel ◽

Fuel Cycle ◽

Nuclear Fuel Cycle ◽

The United States ◽

Spent Fuel ◽

Dry Storage ◽

The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

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ION CHANNELING AND PIXE STUDIES OF S IMPLANTATION IN GaAs

International Journal of PIXE ◽

10.1142/s0129083592000142 ◽

1992 ◽

Vol 02 (02) ◽

pp. 151-159

Author(s):

LIU SHIJIE ◽

WANG JIANG ◽

HU ZAOHUEI ◽

XIA ZHONGHUONG ◽

GAO ZHIGIANG ◽

...

Keyword(s):

Room Temperature ◽

Dislocation Loops ◽

Good Recovery ◽

Electrical Measurements ◽

X Ray ◽

Defect Complexes ◽

Ion Channeling ◽

Transmission Electron ◽

Activation Efficiency ◽

Very High

GaAs (100) crystals were implanted with 100 keV S+ to a dose of 3×1015 cm−2 in a nonchanneling direction at room temperature, and treated with rapid thermal annealing (RTA). He+ Rutherford backscattering and particle-induced X-ray emission in channeling mode in combination with transmission electron microscopy (TEM) were used to study the damage and the lattice location of S atoms. It is revealed that the RTA at 950 °C for 10 sec has resulted in a very good recovery of crystallinity with a few residual defects in the form of dislocation loops, and a very high substitutionality (~90%). The activation efficiency and the Hall mobility of the implanted samples are found to be low after the electrical measurements. Based on these results an extended dopant diffusion effect for the residual defects and a correlation between the electrical properties and defect complexes are suggested.

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Lignin distribution in waterlogged archaeological Picea abies (L.) Karst degraded by erosion bacteria

Holzforschung ◽

10.1515/hf-2013-0228 ◽

2014 ◽

Vol 68 (7) ◽

pp. 791-798 ◽

Cited By ~ 9

Author(s):

Nanna Bjerregaard Pedersen ◽

Uwe Schmitt ◽

Gerald Koch ◽

Claus Felby ◽

Lisbeth Garbrecht Thygesen

Keyword(s):

Picea Abies ◽

Lignin Content ◽

Middle Lamella ◽

Uv Absorbance ◽

Tem Analysis ◽

Lignin Distribution ◽

Transmission Electron ◽

Residual Material ◽

Ray Parenchyma Cells ◽

Very High

Abstract The lignin distribution in poles of waterlogged archaeological Picea abies (L.) Karst, which was decayed by erosion bacteria (EB) under anoxic conditions for approximately 400 years, was topochemically identified by transmission electron microscopy (TEM) and high resolution UV-microspectrophotometry (UMSP). Lignin rich cell wall compartments such as cell corner (CC), compound middle lamella (CML), torus, initial pit border and mild compression wood (CW) appeared morphologically well preserved together with S1 and S3 layers and epithelial and ray parenchyma cells. Residual material (RM) from degraded S2 showed a varied lignin distribution as evidenced by the different local UV-absorbance intensities. However, evaluation of UV-absorbance line spectra of RM revealed no change in conjugation of the aromatic ring system. Presence of RM with both very low and very high lignin absorbances showed evidence for disassembly of lignin during degradation combined with aggregation of lignin fragments and physical movement of these fractions. In contrast to TEM analysis, locally decreasing lignin content was found by UMSP in CML regions.

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