Relevant bibliographies by topics / Nuclear fuel surrogate material

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Nuclear fuel surrogate material'

Author: Grafiati
Published: 4 June 2021
Last updated: 17 February 2022

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Journal articles on the topic "Nuclear fuel surrogate material"

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Park, Seunghoon, Sungyeop Joung, and Jerry Park. "Nuclear Fuel Assay through analysis of Uranium L-shell by Hybrid L-edge/XRF Densitometer using a Surrogate Material." EPJ Web of Conferences 170 (2018): 07009. http://dx.doi.org/10.1051/epjconf/201817007009.

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Assay of L-series of nuclear material solution is useful for determination of amount of nuclear materials and ratio of minor actinide in the materials. The hybrid system of energy dispersive X-ray absorption edge spectrometry, i.e. L-edge densitometry, and X-ray fluorescence spectrometry is one of the analysis methods. The hybrid L-edge/XRF densitometer can be a promising candidate for a portable and compact equipment due to advantage of using low energy X-ray beams without heavy shielding systems and liquid nitrogen cooling compared to hybrid K-edge/XRF densitometer. A prototype of the equipment was evaluated for feasibility of the nuclear material assay using a surrogate material (lead) to avoid radiation effects from nuclear materials. The uncertainty of L-edge and XRF characteristics of the sample material and volume effects was discussed in the article.
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Park, Seunghoon, Sungyeop Joung, and Jerry Park. "Nuclear Fuel Assay through analysis of Uranium L-shell by Hybrid L-edge/XRF Densitometer using a Surrogate Material." EPJ Web of Conferences 170 (2018): 08007. http://dx.doi.org/10.1051/epjconf/201817008007.

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Assay of L-series of nuclear material solution is useful for determination of amount of nuclear materials and ratio of minor actinide in the materials. The hybrid system of energy dispersive X-ray absorption edge spectrometry, i.e. L-edge densitometry, and X-ray fluorescence spectrometry is one of the analysis methods. The hybrid L-edge/XRF densitometer can be a promising candidate for a portable and compact equipment due to advantage of using low energy X-ray beams without heavy shielding systems and liquid nitrogen cooling compared to hybrid K-edge/XRF densitometer. A prototype of the equipment was evaluated for feasibility ofthe nuclear material assay using a surrogate material (lead) to avoid radiation effects from nuclear materials. The uncertainty of L-edge and XRF characteristics of the sample material and volume effects was discussed in the article.
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Devanathan, Ram. "Molecular Dynamics Simulation of Fission Fragment Damage in Nuclear Fuel and Surrogate Material." MRS Advances 2, no. 21-22 (2017): 1225–30. http://dx.doi.org/10.1557/adv.2017.9.

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ABSTRACTWe have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO2) for energy deposition per unit length of 3.9 keV/nm. We did not observe amorphization. The damage mainly consisted of isolated point defects. Only about 1% of the displacements occur on the uranium sublattice. Oxygen Frenkel pairs are an order of magnitude more numerous than uranium Frenkel pairs in the primary damage state. In contrast, previous results show that the ratio of Frenkel pairs on the two sublattices is close to the stoichiometric ratio in ceria. These differences in the primary damage state may lead to differences in radiation response of UO2 and CeO2.
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Li, Xiang, Ya Ting Yang, Cao Fei Fu, Qun Ying Huang, Liu Si Sheng, Zhen Qi Chang, and Christophe C. Serra. "A Microfluidic-Assisted Fabrication of Size-Controlled Porose CeO2 Microspheres as an Analog Production of Nuclear Fuel Beads." Advances in Science and Technology 94 (October 2014): 55–68. http://dx.doi.org/10.4028/www.scientific.net/ast.94.55.

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Porosity-controlled nuclear fuel microsphere is an essential material of fabricating minor actinide-bearing dispersion-type nuclear fuel with the infiltration processes. In this paper, monodisperse and size-controlled spherical oxide nuclear fuel particles with size range of 20μm to 800μm were fabricated by means of microfluidic technology combined with sol-gel process using cerium as a surrogate for plutonium. The porous CeO2 beads with the density range of 25% to 93% T.D. were successfully prepared by the addition of polyethylene glycol 6000 used as a porogen to the feed broth. The uniform U3O8 beads were also prepared at the same experimental conditions as CeO2 beads prepared, which shows the feasibility of the method for fabricating size-controlled monodisperse nuclear fuel beads.
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Jinno, D., Ashwani K. Gupta, and K. Yoshikawa. "Determination of Chemical Kinetic Parameters of Surrogate Solid Wastes." Journal of Engineering for Gas Turbines and Power 126, no. 4 (October 1, 2004): 685–92. http://dx.doi.org/10.1115/1.1772407.

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Results on the thermal decomposition behavior of several important components in solid wastes are presented under controlled chemical and thermal environments. Thermogravimetry (TGA) tests were conducted on the decomposition of cellulose, polyethylene, polypropylene, polystyrene and polyvinyl chloride in inert (nitrogen), and oxidative (air) atmospheres. Inert condition tests were performed at heating rates of 5, 10, 30, and 50°C/min while the oxidative condition tests were performed at one heating rate of 5°C/min. Differential scanning calorimetry (DSC) was also used to measure the heat flow into and out of the sample during thermal decomposition of the material. The TGA results on the mass evolution of the materials studied as a function of temperature showed that the cellulose contained a small amount of moisture whereas no moisture was found in the other materials examined. The DSC curve showed the heat flow into and out of the sample during the process of pyrolysis and oxidative pyrolysis. The temperature dependence and mass loss characteristics of materials were used to evaluate the Arrhenius kinetic parameters. The surrounding chemical environment, heating rate, and material composition and properties affect the overall decomposition rates under defined conditions. The composition of these materials was found to have a significant effect on the thermal decomposition behavior. Experimental results show that decomposition process shifts to higher temperatures at higher heating rates as a result of the competing effects of heat and mass transfer to the material. The results on the Arrhenius chemical kinetic parameters and heat of pyrolysis obtained from the thermal decomposition of the sample materials showed that different components in the waste have considerably different features. The thermal decomposition temperature, heat evolved and the kinetics parameters are significantly different various waste components examined. The amount of thermal energy required to destruct a waste material is only a small faction of the energy evolved from the material. These results assist in the design and development of advanced thermal destruction systems.
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Hayes, John R., and Andrew P. Grosvenor. "Investigation of NdxY0.25–xZr0.75O1.88 inert matrix fuel materials made by a co-precipitation synthetic route." Canadian Journal of Chemistry 94, no. 3 (March 2016): 198–210. http://dx.doi.org/10.1139/cjc-2015-0485.

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Yttria-stabilized zirconia (YSZ) is a material that is being considered for use as an inert matrix fuel in nuclear reactors, but a complete characterization of these materials is required for them to be licensed for use. A series of NdxY0.25–xZr0.75O1.88 materials have been synthesized using a co-precipitation method, and the thermal stability of these materials has been studied by annealing them at 1400 and 1500 °C. (Nd was used as surrogate for Am.) The long-range and local structures of the materials were characterized via powder X-ray diffraction, scanning electron microscopy, wavelength dispersive spectroscopy, and X-ray absorption spectroscopy at the Zr K- and Y K-edges. These results were compared with the previous characterization of Nd-YSZ materials synthesized using a ceramic method. The results indicated that the ordering in the local metal–oxygen polyhedral remains relatively unaffected by the synthetic method, but there was increased long-range disorder in the materials prepared by the co-precipitation method. Further, it was found that the materials produced by the co-precipitation method were unexpectedly unstable when annealed at high temperature. This study highlights the importance of determining the effect of synthetic method on material properties and demonstrates how the co-precipitation route could be used to produce inert matrix fuels.
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Nakanishi, Ryuzo, Morihisa Saeki, Ikuo Wakaida, and Hironori Ohba. "Detection of Gadolinium in Surrogate Nuclear Fuel Debris Using Fiber-Optic Laser-Induced Breakdown Spectroscopy under Gamma Irradiation." Applied Sciences 10, no. 24 (December 16, 2020): 8985. http://dx.doi.org/10.3390/app10248985.

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Fiber-optic laser-induced breakdown spectroscopy (FO-LIBS) was applied to a qualitative and quantitative analysis of gadolinium (Gd) in mixed oxide samples, simulating nuclear fuel debris in the damaged reactors of the Fukushima Daiichi Nuclear Power Station. The surrogate debris was prepared from mixed oxide materials containing Gd2O3, with varying Gd concentrations. The emission spectra of the surrogate debris show that the optical emission lines at 501.5 nm and 510.3 nm are suitable for Gd detection in the nuclear fuel debris. LIBS measurements were further performed under gamma irradiation (0–10 kGy/h), resulting in a decrease in spectral intensities due to radiation-induced damage to the optical fiber. For quantification of Gd, robust calibration curves against gamma irradiation were established from the intensity ratio of Gd (501.5 nm)/Ce (474.5 nm) emission lines, yielding the limits of detection for Gd in the range of 0.03–0.08 wt%. These results demonstrate that FO-LIBS is a potential tool for in situ and remote analysis of nuclear fuel debris.
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Harrison, R. W., R. N. Worth, J. Buckley, and T. Abram. "Atomistic level study of Ce3Si2 oxidation as an accident tolerant nuclear fuel surrogate." Corrosion Science 164 (March 2020): 108332. http://dx.doi.org/10.1016/j.corsci.2019.108332.

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Egeland, G. W., L. D. Zuck, W. R. Cannon, P. A. Lessing, and P. G. Medvedev. "Dry bag isostatic pressing for improved green strength of surrogate nuclear fuel pellets." Journal of Nuclear Materials 406, no. 2 (November 2010): 205–11. http://dx.doi.org/10.1016/j.jnucmat.2010.08.022.

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Firouzdor, Vahid, Lucas Wilson, Kumar Sridharan, Brandon Semerau, Benjamin Hauch, Jamieson Brechtl, James I. Cole, and Todd R. Allen. "Development of Diffusion Barrier Coatings for Mitigation of Fuel-Cladding Chemical Interactions." Key Engineering Materials 507 (March 2012): 3–7. http://dx.doi.org/10.4028/www.scientific.net/kem.507.3.

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Fuel Cladding Chemical Interactions (FCCI) in a nuclear reactor occur due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and the formation of low melting point eutectic compounds. Deposition of diffusion barrier coatings of a thin oxide on the inner surface of the cladding can potentially reduce or delay the onset of FCCI. This study examines the feasibility of using nanofluid-based electrophoretic deposition (EPD) process to deposit coatings of titanium oxide, yttria-stabilized zirconia (YSZ) and vanadium oxide. The deposition parameters, including the nanofluid composition, current, and voltage were optimized for each coating material using test flat substrates of T91 ferritic-martensitic steel. Diffusion characteristics of the coatings were investigated by diffusion couple experiments using the fuel surrogate cerium. These diffusion couple studies performed in the temperature range of 560°C and 585°C showed that the oxide coatings significantly reduce the solid state inter-diffusion between cerium to steel.
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Dissertations / Theses on the topic "Nuclear fuel surrogate material"

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Roleček, Jakub. "Příprava keramických materiálů se zvýšenou tepelnou vodivostí pro jaderné aplikace." Master's thesis, Vysoké učení technické v Brně. Fakulta strojního inženýrství, 2014. http://www.nusl.cz/ntk/nusl-231722.

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Oxid uraničitý (UO2) je v současnosti nejčastěji používaným materiálem jakožto palivo v komerčních jaderných reaktorech. Největší nevýhodou UO2 je jeho velmi nízká tepelná vodivost, a protože se při štěpení UO2 v jaderném reaktoru vytváří velké množství tepla, vzniká v UO2 peletě velký teplotní gradient. Tento teplotní gradient způsobuje vznik velkého tepelného napětí uvnitř pelety, což následně vede k tvorbě trhlin. Tyto trhliny napomáhají k šíření štěpných plynů při vysoké míře vyhoření paliva. Tvorba trhlin a zvýšený vývin štěpného plynu posléze vede ke značnému snížení odolnosti jaderného paliva. Tato práce se zabývá problematikou zvyšování tepelné vodivosti jaderného paliva na modelu materiálu (CeO2). V této práci jsou studovány podobnosti chování CeO2 a UO2 při konvenčním slinováním a při „spark plasma sintering.“ Způsob jak zvýšit tepelnou vodivost použitý v této práci je včlenění vysoce tepelně vodivého materiálu, karbidu křemíku (SiC), do struktury CeO2 pelet. Od karbidu křemíku je očekáváno, že zvýší tok tepla z jádra pelety, a tím zvýší tepelnou vodivost CeO2. V této práci je také porovnávána podobnost chování SiC v CeO2 matrici s chováním SiC v UO2, které bylo popsáno v literatuře.
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Krivtchik, Guillaume. "Analysis of uncertainty propagation in nuclear fuel cycle scenarios." Thesis, Grenoble, 2014. http://www.theses.fr/2014GRENI050/document.

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Les études des scénarios électronucléaires modélisent le fonctionnement d’un parcnucléaire sur une période de temps donnée. Elles permettent la comparaison de différentesoptions d’évolution du parc nucléaire et de gestion des matières du cycle, depuis l’extraction duminerai jusqu’au stockage ultime des déchets, en se basant sur des critères tels que les puis-sances installées par filière, les inventaires et les flux, en cycle et aux déchets. Les incertitudessur les données nucléaires et les hypothèses de scénarios (caractéristiques des combustibles, desréacteurs et des usines) se propagent le long des chaînes isotopiques lors des calculs d’évolutionet au cours de l’historique du scénario, limitant la précision des résultats obtenus. L’objetdu présent travail est de développer, implémenter et utiliser une méthodologie stochastiquede propagation d’incertitudes dans les études de scénario. La méthode retenue repose sur ledéveloppement de métamodèles de calculs d’irradiation, permettant de diminuer le temps decalcul des études de scénarios et de prendre en compte des perturbations des paramètres ducalcul, et la fabrication de modèles d’équivalence permettant de tenir compte des perturbationsdes sections efficaces lors du calcul de teneur du combustible neuf. La méthodologie de calculde propagation d’incertitudes est ensuite appliquée à différents scénarios électronucléairesd’intérêt, considérant différentes options d’évolution du parc REP français avec le déploiementde RNR
Nuclear scenario studies model nuclear fleet over a given period. They enablethe comparison of different options for the reactor fleet evolution, and the management ofthe future fuel cycle materials, from mining to disposal, based on criteria such as installedcapacity per reactor technology, mass inventories and flows, in the fuel cycle and in the waste.Uncertainties associated with nuclear data and scenario parameters (fuel, reactors and facilitiescharacteristics) propagate along the isotopic chains in depletion calculations, and throughoutthe scenario history, which reduces the precision of the results. The aim of this work isto develop, implement and use a stochastic uncertainty propagation methodology adaptedto scenario studies. The method chosen is based on development of depletion computationsurrogate models, which reduce the scenario studies computation time, and whose parametersinclude perturbations of the depletion model; and fabrication of equivalence model which takeinto account cross-sections perturbations for computation of fresh fuel enrichment. Then theuncertainty propagation methodology is applied to different scenarios of interest, consideringdifferent options of evolution for the French PWR fleet with SFR deployment
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Turner, Joel David. "The performance of a nuclear fuel-matrix material in a sealed CO₂ system." Thesis, University of Manchester, 2013. https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html.

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An advanced concept high temperature reactor (HTR) design has been proposed - The ‘U-Battery’, which utilises a unique sealed coolant loop, and is intended to operate with minimal human oversight. In order to reduce the need for moving parts within the design, CO2 has been selected as a candidate coolant, potentially allowing a naturally circulated system. HTR fuel is held within a semi-graphitic fuel-matrix material, and this has not previously been tested within a CO2 environment. Graphite in CO2 is subject to two oxidation reactions, one thermally driven and one radiolytically. As such, the oxidation performance of fuel-matrix material has been tested within CO2 at both high temperatures and under ionising radiation within a sealed-system. Performance has been compared to that of the Gilsocarbon and NBG-18 nuclear graphite grades. Gilsocarbon is the primary graphite grade used within the currently operating AGR fleet within the UK, and as such is known to have acceptable oxidation performance under reactor conditions. NBG-18 is a modern graphite grade, and is a candidate material for use within the U-Battery. Virgin characterisation of all materials was performed, including measurements of bulk mass and volume, skeletal volumes and surface areas. High-resolution optical microscopy has also been performed and pore size distributions inferred from digital image analysis. All results were seen to agree well with literature values, and the variation between samples has been quanti- fied and found to be < 10% between samples of Gilsocarbon, and < 4% for samples of fuel-matrix and NBG-18. Thermal performance of fuel-matrix material was observed between 600 °C – 1200 °C and seen to be broadly comparable to that of the nuclear graphite grades tested. NBG-18 showed surprisingly poor performance at 600°C, with an oxidation rate of 3×10−4%/min, approximately ten times faster than Gilsocarbon in similar conditions, and three times faster than fuel-matrix material. The radiolytic oxidation performance of fuel-matrix material and NBG-18 has been observed by irradiating sealed quartz ampoules. Ampoules were pressurised with CO2 prior to irradiation, and the pressure after 30 days of irradiation was measured and seen to fall by 50%. Radiolytic oxidation, and the subsequent radiolysis of the reaction product, CO, was seen to cause significant carbonaceous deposition on the internal surfaces of the ampoule and throughout the samples. Due to the short irradiation times available in the present study, an investigation of the microporosity within irradiated samples has been carried out, using nitrogen adsorption and small-angle neutron scattering (SANS). Pore size distributions produced from SANS show the closure of microporosity within NBG-18, most likely as a result of low-temperature neutron irradiation.As a result of this work, CO2 is no longer a candidate coolant for use with the U-Battery design, due to the rapid deposition observed following irradiation.
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MOREIRA, JOSE P. "Avaliacao de sistemas de controle e contabilidade de material nuclear nas operacoes de conversao de uranio." reponame:Repositório Institucional do IPEN, 1994. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10414.

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Made available in DSpace on 2014-10-09T14:04:58Z (GMT). No. of bitstreams: 1 05832.pdf: 3575221 bytes, checksum: 15057ef2624309cb63dafed6754a1cfe (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Lisowski, Eva. "Evaluation of material attractiveness to non-state actors of various nuclear materials in Thorium fuel cycles." Thesis, Massachusetts Institute of Technology, 2020. https://hdl.handle.net/1721.1/127306.

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Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, May, 2020
Cataloged from the official PDF of thesis.
Includes bibliographical references (pages 42-43).
Thorium-based fuel cycles for advanced nuclear reactors have been explored to utilize thorium resources in nations where uranium is scarce, increase fissile material utilization, and enhance proliferation resistance. As a stepping stone, thorium-based fuels have been paired with pressure tube heavy water reactors because of their high neutron economy and online refueling capability. However, thorium fuel cycles have raised proliferation concerns regarding the presence of U-233 following the irradiation of fuel bundles. The presence of Pa-233, which decays into pure U-233, and the creation of Pu-239 due to the neutron capture of U-238 in mixed lightly-enriched uranium (LEU)/Thorium fuels, are also causes for proliferation concern.
Based on a method developed in a previous study, the material attractiveness to non-state actors of fissile materials present in a 40%LEU/60%Th fuel lattice concept was evaluated for six metrics: bare critical mass (BCM), heat content, net weight, acquisition time, dose rate, and processing time & complexity. The lattice, composed of 35 fuel pins and a central ZrO2 displacer rod, was modeled and depleted in the OpenMC reactor physics software, over a range of burnups up to 40 MWd/kg followed by two years of cooling. It was found that the material attractiveness of uranium isotopes in the irradiated fuel bundle was Very Low due to the high fraction of U-238 present in the fuel and the assumed lack of enrichment capabilities among non-state actors. However, for a state with basic enrichment capabilities, this fuel may be attractive. The attractiveness of plutonium isotopes was also found, as expected in a thorium-cycle, to be Very Low.
However, the low BCM and heat content of this mixture reveals that it could be attractive to states that can easily acquire the material and do not need to rely on the theft of many fuel bundles to acquire an IAEA Category I quantity of material. Further investigation of the material attractiveness to states is required. Material attractiveness evaluations are important to informing future decisions regarding which fuel bundles to select when designing advanced reactor facilities and developing methods to safeguard them.
by Eva Lisowski.
S.B.
S.B. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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Shooshpasha, Issa. "Performance of clay based buffer material developed for use in a nuclear fuel waste disposal vault." Thesis, McGill University, 1992. http://digitool.Library.McGill.CA:80/R/?func=dbin-jump-full&object_id=26016.

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This study is designed to evaluate water diffusion parameters of clay based buffer material (mixture of clay and sand) used in nuclear fuel waste disposal vault for the following three cases: (a) No volume change in the soil fabric with no air loss. (b) No volume change in the soil fabric with air loss. (c) A complete change in the soil fabric with air loss.
One dimensional diffusion type unsaturated flow equation was solved by finite difference method. Powell's optimization technique was used to minimize the material parameters in the proposed diffusion function. The adopted technique makes the use of both theory and experimental data. In this concern several tests have been performed for the three aforementioned conditions to measure the volumetric water content and the soil water potential distributions as a function of time and space.
The calculated diffusion parameters were used to predict the volumetric water content and the soil water potential as a function of space and time for longer period of time.
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Ferry, Sara Elizabeth. "Selection and implementation of deterministic and probabilistic models for the prediction of stress corrosion cracking in used nuclear fuel containment material." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/106766.

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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.
"June 2016." Cataloged from PDF version of thesis.
Includes bibliographical references (pages 391-408).
Stainless steel canisters encased in vented concrete overpacks are used to store used nuclear fuel at interim spent fuel storage installations in the United States. These storage systems are exposed to the elements. There is concern that, over time, a deliquesced salt film could develop on the stainless steel canister. Such a film can create a corrosive environment in stainless steels. If a stress is present on or in then material, stress corrosion cracking (SCC) could be a possibility. Because the canister welds are not stress-relieved in order to avoid sensitization of the steel, residual stresses are expected to be present in the canisters. Thus, there is interest in determining the likelihood that (a) a sufficiently corrosive film develops on a stainless steel used fuel canister (b) there are sufficiently high stresses in the material at the location of the corrosive film (c) SCC initiates and (d) the crack propagates through the canister wall, resulting in canister failure. This thesis begins with the assumption that a corrosive film has developed on the canister surface, and pits have begun to initiate. It investigates various methods of modeling SCC in the canister wall after the point of corrosion pit initiation. An extensive literature review was carried out in order to understand the different SCC models that currently exist in the literature. A figure-of-merit was developed to decide which models were the most likely to be helpful to the modeling of SCC in used fuel canisters. The figure-of-merit was then used to select the most promising models. These models were then used to write MATLAB@ simulations that could be used to predict time-to-failure in canisters due to SCC once corrosion pits have begun to grow. The results of these simulations are then considered and compared, and used to inform recommendations for future development of a useful predictive model of SCC in used fuel canisters.
by Sara Elizabeth Ferry.
S.M.
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Denis, Julien. "Dynamic modelling of local fuel desorption and retention in the wall of nuclear fusion reactors for auto-consistent plasma-wall interaction simulations." Electronic Thesis or Diss., Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0682.

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Dans les réacteurs à fusion nucléaire, un échange constant de particules se développe entre le plasma et la paroi. Ce phénomène, appelé recyclage, présente un intérêt crucial pour ces réacteurs car, une fois le plasma établi, le recyclage représente la principale source de particules pour le plasma. Une connaissance complète du recyclage est donc essentielle pour assurer un contrôle fiable de la densité plasma ainsi qu'une performance optimale du réacteur. Cependant, les études in-situ du recyclage demeurent complexes. La modélisation numérique peut assister dans la compréhension de ce phénomène. Les codes de transport de plasma de bord présentent la description la plus avancée de l’interaction plasma-paroi. La description du recyclage reste néanmoins partielle : la réflexion atomique est déjà traitée tandis que la désorption moléculaire est fixée ad-hoc par l’utilisateur. Cette thèse de doctorat se concentre sur le développement d’une extension au code de transport SolEdge2D-EIRENE, dénommée D-WEE, dont l’objectif est de modéliser la dynamique de la désorption. Afin d’initialiser D-WEE, une séquence de décharges plasma est simulée pour modéliser l’opération d’un tokamak. La dynamique simulée pendant ces décharges est étudiée, révélant des comportements intéressants qui pourraient impacter l’opération du réacteur. Pour évaluer la pertinence de la simulation, une confrontation avec l'expérience est effectuée et révèle un accord qualitatif entre la chute de pression post-décharge simulée (avec une tendance en t^{-0.8}) et celle observée expérimentalement. Le taux de rétention simulé pendant la décharge est étudié et présente un accord qualitatif avec l'expérience
In all plasma devices, a constant exchange of particle develops between the plasma and the wall. This phenomenon, referred to as recycling, is of critical interest for those reactors as, once the plasma is established, particle recycling represents the main particle source for the plasma. A complete understanding of the recycling phenomenon is therefore essential to ensure a reliable plasma density control and optimum performance. However, the in-situ experimental study of recycling remains challenging. Modelling can assist in the understanding of this phenomenon.The edge-plasma transport codes present the most-advanced description of the plasma-wall interaction. However, the description of recycling remains partial: atomic reflection is already handled while molecular desorption is set ad-hoc by the code user. This PhD focuses on the development of an extension of the SolEdge2D-EIRENE transport code, named D-WEE, whose goal is to model the dynamics of desorption.To initialise D-WEE, a sequence of plasma discharges is simulated with D-WEE to model a tokamak operation. The simulated wall dynamics during those discharges is studied, reveling some interesting behaviours that could impact the reactor operation. To assess the relevance of the simulated wall dynamics, a confrontation to post-pulse experimental pressure measurement is performed which reveals a qualitative agreement between the temporal pressure drop obtained in the simulation (with a t^{-0.8} trend) and the one observed experimentally under ILW configuration. The retention rate during the discharge is also studied and reproduces qualitatively the experimental trends
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Lage, Aldo Márcio Fonseca. "Modelagem geométrica computacional das etapas de prensagem e sinterização de pastilhas e de laminação de placas combustíveis em dispersão de microesferas de (Th,25%U) O2 em matriz de aço inoxidável." CNEN - Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, 2005. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=46.

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Neste trabalho foi realizada a modelagem geométrica computacional das Cetapas de prensagem e sinterização da pastilha e da laminação da placa de combustível nuclear contendo microesferas de (Th,25%U)O2 dispersas em matriz de aço inoxidável com o objetivo de avaliar a distribuição destas microesferas nas diversas etapas do processamento. As regras de modelagem foram desenvolvidas baseadas nos parâmetros de cada etapa da fabricação da placa combustível. Para isto foram obtidas placas através do processamento por laminação de molduras de chapas de aço inoxidável, contendo pastilha fabricadas com microesferas de (Th,25%U)O2 com carregamentos de 10, 20 e 40% em peso de combustível disperso em matriz de aço inoxidável. Os dados das placas com carregamentos de 30 e 50% foram obtidos por interpolação da curva. As microesferas, obtidas pelo processo sol-gel, foram previamente secas, reduzidas e sinterizadas a 1700oC, durante 2 horas, sob atmosferas de hidrogênio. As microesferas sinterizadas alcançaram uma densidade de cerca de 98% da densidade teórica, e possuem um diâmetro médio de cerca de 300 mm e uma elevada resistência à fratura, de aproximadamente 40 N/microesfera. As regras implementadas neste modelo foram aplicadas nas coordenadas dos centros das esferas virtuais, que simulam as microesferas combustíveis de (Th,25%U)O2, obtendo-se novas coordenadas espaciais para cada uma delas nas etapas de prensagem e sinterização da pastilha e da laminação da placa combustível. Este modelo foi projetado com o uso de técnicas de análise de sistema estruturada, implementado utilizando a linguagem de programação Delphi e os resultados visualizados através do programa AutoCAD. Os resultados do modelo foram validados comparando-se as frações volumétricas experimentais em cada um dos carregamentos estudados com as frações simuladas. Este trabalho será de grande valia para o estudo do carregamento de microesferas na placa combustível, permitindo obter um combustível de elevado desempenho mecânico, térmico e neutrônico mesmo em mais alto carregamento.
The computational geometric modeling of the pressing, sintering and lamination stages for nuclear fuel plates composed by (Th,25%U)O2, microspheres dispersed into stainless steel matrix has been done in order to investigate the microspheres distribution in the various processing stages. The modeling standards were based on the parameters related to each fuel plate manufacturing stage. Accordingly, the plates were obtained through lamination processing of stainless steel plate frames comprising (Th,25%U)O2 microspheres pellets dispersed into stainless steel powder with loading of 10, 20 and 40% of microspheres dispersed into stainless steel matrix. The data for plates with loading of 30 and 50% have been obtained by linear interpolation. The microspheres produced by the sol-gel method were previously reduced and sintered at 1700 0C during 2 hours at hydrogen atmosphere. These sintered microspheres have reached about 98% of the theoretical density, with a mean diameter of 300 mm and a high resistance to fracture, near to 40 N/microsphere. The implemented standards in this model were applied at the virtual spheres center coordinates, which simulate the (Th,25%U)O2 fuel microspheres, and generate the new spatial coordinates to each of them in the pressing, sintering and lamination stages. This model was developed using structured system analysis techniques and it has been implemented using the Delphi programming language. The results were displayed through the AutoCAD program and validated comparing the experimental volumetric fractions in each of the studied loading, with the simulated fractions. The results indicate that this work could be a powerful tool in the investigation of microspheres loading in the fuel plate, allowing the attainment of a high mechanical and neutronic performance fuel, even for higher level loading.
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10

Tyrpekl, Vaclav. "Effet matériaux lors de l'interaction corium-eau : analyse structurale des débris d'une explosition vapeur et mécanismes de solidification." Phd thesis, Université de Strasbourg, 2012. http://tel.archives-ouvertes.fr/tel-00758983.

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Ce travail a été réalisé en cotutelle entre l'Université Charles à Prague (République Tchèque) et l'Université de Strasbourg (France). Il a également profité d'une coopération entre l'Institut de Chimie Inorganique de l'Académie des Sciences de République Tchèque et le Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, Cadarache, France). Les résultats des travaux ont contribué au projet OCDE / AEN Serena 2 (Programme portant sur l'étude des effets d'une explosion de vapeur dans un réacteur nucléaire à eau). La thèse présentée se situe dans le domaine de la sûreté nucléaire et de la science des matériaux. Elle traite de l'Interaction Combustible-Réfrigérant (ICR, ou FCI en anglais pour Fuel-Coolant Interaction) susceptible d'intervenir lors d'un accident grave de réacteur nucléaire et actuellement à l'étude dans les programme de R&D. Au cours d'un accident de fusion d'un coeur de réacteur, les matériaux fondus peuvent interagir avec le liquide de refroidissement (eau légère), aussi appelé réfrigérant. Cette interaction peut se produire à l'intérieur de la cuve ou, en cas de rupture de celle-ci, à l'extérieur. Ces deux scénarios sont couramment appelés Interaction Combustible-Réfrigérant en- et hors- cuve et se distinguent de par les conditions du réacteur lors de l'accident : pression du système, degré de sous refroidissement de l'eau, etc. L'interaction entre le combustible fondu et le liquide de refroidissement peut évoluer vers une détonation thermique appelée "explosion de vapeur" qui peut endommager le réacteur, voire compromettre l'intégrité du confinement. Des expériences récentes ont montré que la composition du combustible a un effet majeur sur l'apparition et le rendement d'une telle explosion. En particulier, des comportements différents ont été observés entre un matériau simulant, l'alumine, qui explose très facilement, et diverses compositions de corium prototypique (80 m. % UO2, 20% m.% ZrO2). Cet "effet matériau" a suscité un intérêt nouveau pour les analyses post-expériences des débris issus de l'ICR afin de déterminer les mécanismes qui interviennent au cours de ces phénomènes extrêmement rapides. La thèse est organisée en neuf chapitres. Le chapitre 1 constitue une introduction générale et présente le contexte d'un accident grave d'un réacteur nucléaire. Quelques exemples d'accidents graves (Three Miles Island 1979, Tchernobyl 1986 et Fukushima 2011) sont brièvement abordés. Le chapitre 2 résume les aspects théoriques de l'interaction combustible-réfrigérant. Il est divisé en quatre parties correspondant aux quatre étapes généralement rencontrées lors du mécanisme d'ICR i) Prémélange - le combustible fondu, versé dans l'eau, se fragmente en gouttelettes grossières qui s'isolent d'un film de vapeur. ii) Déclenchement - le film de vapeur entourant les gouttes de combustible est déstabilisé, permettant ainsi la fragmentation fine du combustible. iii) Propagation - la fragmentation du combustible se propage à l'ensemble du prémélange, augmentant ainsi la surface de contact entre le combustible fondu et l'eau. Ceci conduit à une production intense de vapeur à grande échelle. iv) Expansion (explosion) - l'énergie thermique transférée du combustible à l'eau est transformée en travail mécanique de la vapeur.[...]
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Books on the topic "Nuclear fuel surrogate material"

1

Commission, Canadian Nuclear Safety. Transportation security plans for category I, II or III nuclear material. [Ottawa]: Canadian Nuclear Safety Commission, 2003.

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2

Brown, C. Physical protection requirements for categories I, II, and III material at fuel cycle facilities. Washington, DC: Division of Fuel Cycle Safety and Safequards, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 1999.

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IAEA Technical Meeting on Fissile Material Management Strategies for Sustainable Nuclear Energy (2005 Vienna, Austria). Fissile material management strategies for sustainable nuclear energy: Proceedings of an IAEA Technical Meeting on Fissile Material Management Strategies for Sustainable Nuclear Energy, held in Vienna, 12-15 September 2005. Vienna, Austria: International Atomic Energy Agency, 2007.

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McGuire, Stephen A. A regulatory analysis on emergency preparedness for fuel cycle and other radioactive material licensees: Final report. Washington, D.C: Division of Accident Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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McGuire, Stephen A. A regulatory analysis on emergency preparedness for fuel cycle and other radioactive material licensees: Draft report for comment. Washington, D.C: Division of Risk Analysis and Operations, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1985.

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International Atomic Energy Agency; IAEA. Measures to prevent, intercept and respond to illicit uses of nuclear material and radioactive sources: proceedings: International conference held in Stockholm, Sweden, 7-11 May 2001. Vienna: IAEA, 2002.

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Dwyer, P. Use of perimeter alarms at fuel fabrication facilities using or possessing formula quantities of strategic special nuclear material. Washington, DC: Division of Safeguards and Transportation, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 1988.

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William, Walker. Fissile material stocks: Characteristics, measures and policy options / William Walker and Frans Berkhout. New York: United Nations, 1999.

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Office, General Accounting. Nuclear nonproliferation: Status of U.S. efforts to improve nuclear material controls in newly independent states. Washington, D.C: The Office, 1996.

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Parvin, Daniel Francis. Investigations using computer modelling into isotope discriminaton [sic] techniques relating to active neutron interrogation of nuclear material within irradiated fuel hulls. Birmingham: University of Birmingham, 1997.

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Book chapters on the topic "Nuclear fuel surrogate material"

1

Wang, Hong, and Jy-An John Wang. "Experimental Study on Surrogate Nuclear Fuel Rods Under Reversed Cyclic Bending." In Fatigue and Fracture Test Planning, Test Data Acquisitions and Analysis, 19–36. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2017. http://dx.doi.org/10.1520/stp159820160051.

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Pshakin, G. M., and A. N. Tuzov. "Security and Safeguards Aspects of Nuclear Material Handling in Russia." In Mixed Oxide Fuel (Mox) Exploitation and Destruction in Power Reactors, 43–49. Dordrecht: Springer Netherlands, 1995. http://dx.doi.org/10.1007/978-94-017-2288-9_5.

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Carbol, P., D. H. Wegen, T. Wiss, and P. Fors. "Spent Fuel as Waste Material." In Comprehensive Nuclear Materials, 389–420. Elsevier, 2012. http://dx.doi.org/10.1016/b978-0-08-056033-5.00106-3.

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Kommalapati, Raghava R., Fiifi Asah-Opoku, Hongbo Du, and Ziaul Huque. "Monte Carlo Simulations of Nuclear Fuel Burnup." In Nuclear Material Performance. InTech, 2016. http://dx.doi.org/10.5772/62572.

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Carbol, Paul, Detlef H. Wegen, Thierry Wiss, Patrik Fors, Christophe Jegou, and Kastriot Spahiu. "Spent Nuclear Fuel as Waste Material." In Comprehensive Nuclear Materials, 347–86. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-803581-8.10374-1.

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URSU, IOAN. "Nuclear Material Recovery from Irradiated Fuel and Recycling." In Physics and Technology of Nuclear Materials, 318–66. Elsevier, 1985. http://dx.doi.org/10.1016/b978-0-08-032601-6.50015-5.

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"Fuel material enrichment for various types of reactors." In Physics and Technology of Nuclear Materials, 463. Elsevier, 1985. http://dx.doi.org/10.1016/b978-0-08-032601-6.50020-9.

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Pope, R. B. "Packaging and transport of radioactive material in the nuclear fuel cycle." In Nuclear Fuel Cycle Science and Engineering, 558–98. Elsevier, 2012. http://dx.doi.org/10.1533/9780857096388.4.558.

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Gao, Yuan, Andreas Enqvist, and James Baciak. "Effect of spent nuclear fuel neutron aging on dry cask shielding material." In Micro and Nanostructured Composite Materials for Neutron Shielding Applications, 401–20. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-819459-1.00015-5.

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Ion, Sue. "Nuclear fission." In Energy... beyond oil. Oxford University Press, 2007. http://dx.doi.org/10.1093/oso/9780199209965.003.0008.

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This chapter will cover the nuclear fission option as a future energy supply, and will essentially address the question: can nuclear fission plug the gap until the potential of nuclear fusion is actually realized? (The potential for fusion is considered in detail chapter 7.) To put this question into context, let us first look at some of the key issues associated with nuclear fission, which currently supplies around one fifth of the UK’s electricity. Most large scale power stations produce electricity by generating steam, which is used to power a turbine. In a nuclear power station, the principle is the same, but instead of burning coal, oil, or gas to turn water into steam, the heat energy comes from a nuclear reactor. A reactor contains nuclear fuel, which remains in place for several months at a time, but over that time it generates a huge amount of energy. The fuel is usually made of uranium, often in the form of small pellets of uranium dioxide, a ceramic, stacked inside hollow metal tubes or fuel rods, which can be anything from a metre to four metres in length, depending on the reactor design. Each rod is about the diameter of a pencil, and the rods are assembled into carefully designed bundles, which in turn are fixed in place securely within the reactor. There are two isotopes (or different types) of uranium, and only one of these is a material which is ‘fissionable’—that is to say, if an atom of this uranium isotope is hit by a neutron, then it can split into two smaller atoms, giving off energy in the process and also emitting more neutrons. This, and other pathways, are illustrated in Fig. 6.1 (Source: CEA). Controlling the reaction, so that the energy from the fission of uranium atoms is given out slowly over a period of years, requires two aspects of the process to be carefully balanced. 1. First, there must be enough fissile atoms in the fuel so that—on average— each fission leads to exactly one other. Any fewer, and the reaction will die away.
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Conference papers on the topic "Nuclear fuel surrogate material"

1

Ozaltun, Hakan. "Effects of Cladding Material on Irradiation Performance of Monolithic Mini-Plates." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60120.

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Monolithic, plate-type fuels are the proposed fuel form for the conversion of the research and test reactors to achieve higher uranium densities within the reactor core. This fuel type is comprised of a low enrichment, a high density U-10Mo alloy fuel-foil, which is sandwiched between diffusion barriers and encapsulated in a cladding material. To understand the irradiation performance, fuel-plates are being benchmarked for large number of parameters. In this work, effects of the cladding material were studied. In particular, a monolithic fuel-plate with U7Mo foil and Zry-4 cladding was simulated to explore feasibility of using Zircaloy as a surrogate cladding material. For this, a selected mini-plate from RERTR-7 tests was simulated first with as-run irradiation history. By using same irradiation parameters, a second case, a plate with U10Mo fuel and Al6061 cladding was simulated to make a comparative assessment. The results indicated that the plate with Zircaloy cladding would operate roughly 50 °C hotter compared with the plate with Aluminum cladding. Larger displacement profiles along the thickness for the plate with Zircaloy cladding were observed. Higher plastic strains occur for the plate with Aluminum cladding. The results have revealed that any pre-irradiation stresses would be relieved relatively fast in reactor and the fuel-foil would be essentially stress-free during irradiation. The fuel stresses however, develop at reactor shutdown. The plate with Zircaloy cladding would have higher residual stresses due to higher pre-shutdown temperatures. Similarly, the stresses magnitudes are higher in the foil core for the plates with Zircaloy cladding. Finally, pressure on the fuel is significantly higher for the plates with Zircaloy cladding. Overall, employing a Zircaloy as surrogate cladding material did not provide a better thermo-mechanical performance compared with the Aluminum cladding.
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Foral, Štěpán, David Salamon, Karel Katovský, Jan Varmuža, and Jakub Roleček. "Influence of Silicone Carbide on the Reactivity of Nuclear Fuels Using Cerium Dioxide as a Surrogate Material." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-31262.

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The most researched material in nuclear power industry is uranium dioxide however due to strict safety and sanitary restrictions this material can be researched only in specialized research institutes and universities which have sufficient technological background. For this reason it can be suitable to find material which would show physical properties similar to UO2 but would not suffer by the strict limitations in storage and handling. In this case much more workplaces could be incorporated in the material research and the list of investigated problems could be significantly enlarged. One of the possible substitutional materials is the cerium dioxide (CeO2) which shows similar chemical and physical properties like UO2 and in some cases shows also similar neutronic properties. The laboratory research was focused on comparing of the basic neutronic properties. For the comparison of nuclear properties the JANIS software [1] was used as it contains cross section libraries of both materials. It will be shown that similarity of both materials is significant and in several cases application of cerium dioxide as alternative material is possible with sufficient accuracy. As an example of the use of CeO2, research of influence of the SiC content on the reactivity of nuclear fuels is presented in this work.
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Louie, David L. Y., Yifeng Wang, Rekha Rao, Alec Kucala, and Jessica Kruichak. "Injectable Sacrificial Material System to Contain Ex-Vessel Molten Corium in Nuclear Accidents." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81440.

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An ongoing Sandia National Laboratories’ (SNL) research study is evaluating a potential design of an injectable sacrificial material (SM) system that could contain and cool corium ejected from a reactor vessel lower head failure during a potential severe accident involving melting fuel at a commercial light water nuclear reactor (LWR). An injectable system could be installed at any existing LWR, without significant modification to the cavity or to the drywell pedestal region of the plant. The conceptual design under consideration is a passive system. The SM is being optimized to quickly cool the corium mixture while creating gas to form porosity in the solid, such that subsequent water flooding can penetrate the structure and provide additional cooling. The SM would form a barrier and limit corium-concrete interactions. This three-year project takes a joint experimental and computational approach. In this paper, we will first discuss the success of our small-scale experiments conducted on the interactions between the surrogate corium material (SCM) and SM, used to evaluate the injectable concept. A larger experimental study, currently underway, will further validate the injectable concept, with a focus on accurately measuring interactions. This paper details the modeling study and its progress, including modeling the experiments on a surrogate system and extending the model to bench-scale corium flow from validation experiments. The project’s modeling studies will use the SNL engineering code suite SIERRA Mechanics to understand the interaction of injectable SM and molten corium and predict corium spreading. Spreading is modeled using a level set method to track the front in conjunction with a pressure-stabilized finite element method on the fully three-dimensional mass, momentum, and energy conservation equations. Using this diffuse-interface method, the corium spreading front can be tracked and an appropriate pseudo-solidification viscosity models can be implemented to accurately model the corium spreading physics. Finally, an injectable SM delivery system is discussed along with its deployment to the six-common commercial LWR designs currently operating in the United States. At the end of this project, a simplified model based on SIERRA simulations will be developed for implementation into MELCOR, a severe reactor analysis code, developed at SNL for the U.S. Nuclear Regulatory Commission. This will allow us to demonstrate the ability of the injectable SM system to mitigate the ex-vessel corium spreading, provide containment and negate the release of radionuclides.
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4

de Haas, Simon, David Chu, Kevin Ellis, Matthew White, Ben Lindley, Peter Smith, Julian Murgatroyd, Andrew Grief, Mike Leddy, and Mike Yule. "Integrated Design of a Reactor Core for the Rolls-Royce Small Modular Reactor Project." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81311.

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Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR. The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material. A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken. This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.
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Ursˇicˇ, Mitja, Matjazˇ Leskovar, Ivo Kljenak, and Borut Mavko. "Modelling of Material Properties Influence in Fuel-Coolant Interaction Codes." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75039.

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The melt droplets crust formation modelling, which is used in current FCI (fuel coolant interaction) codes, is rather basic. In the paper the development of the melt droplet heat transfer model, which enables the treatment of the material properties influence on the steam explosion is presented. The model is enough complex to correctly predict the crust development during the melt droplets cooling in the premixing phase. At the same time the model is enough simple that it can be practically implemented into FCI codes and is thus being an optimal model for FCI applications. Fragmentation criteria are derived in order to take into account the influence of the formed crust on the steam explosion process. The derived criteria are based on experimental results and the thin plate approximation. To enable the use of the model and the fragmentation criteria in FCI codes with Eulerian formulation, adequate transport equations for model parameters are given.
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6

Jiang, Hao, Jy-An John Wang, and Hong Wang. "Potential Impact of Interfacial Bonding Efficiency on Used Nuclear Fuel Vibration Integrity During Normal Transportation." In ASME 2014 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/pvp2014-29067.

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Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet–pellet and pellet–clad interfaces on surrogate of used nuclear fuel (UNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reversible bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, κ, the flexural rigidity EI can be estimated as EI=M/κ. The impacts of interfacial bonding efficiency include the moment carrying capacity distribution between pellets and clad and cohesion influence on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet–pellet and pellet–clad interfaces. FEA models will be further used to study UNF vibration integrity.
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Wang, Hong, Jy-an John Wang, and Hao Jiang. "Fatigue Behavior of Spent Nuclear Fuel Rods in Simulated Transportation Environment." In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-65842.

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Nuclear fuel rod is composed of cladding tube and a specified number of fuel pellets contained. In the United States, spent nuclear fuel (SNF) is expected to be transported to at least one storage facility before permanent disposal. The fatigue behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending must be understood in order to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup SNFs (>45 GWd/MTU). This report presents the experimental results related to Zircaloy (Zry)-4-based surrogate rods and high-burnup SNFs, based on recent work performed at Oak Ridge National Laboratory (ORNL). The surrogate rod was made of Zry-4 cladding and alumina pellets, and high-burnup fuel rods were discharged from H.B. Robinson pressurized water reactor. The reversed cyclic bending testing was conducted at 5 Hz under loading control. The effect of pre-hydriding and burnup or irradiation on the flexural rigidity and fatigue life of cladding-pellet system were discussed. The fatigue data obtained are extremely useful to the future certification of SNF storage and transportation cask.
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Segawa, Tomoomi, Koichi Kawaguchi, Katsunori Ishii, Masahiro Suzuki, Joji Tachihara, Kiyoto Takato, Takatoshi Okita, Hiroshi Satone, and Michitaka Suzuki. "Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16215.

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Abstract To minimize retention of nuclear fuel materials in glove box components and curtail the external exposure dose, plutonium and uranium mixed oxide powder adhesion prevention technology involving nanoparticle coating of the acrylic panels of the glove box is developed. The nanoparticle coating reduces the van der Waals force between alumina particles and the acrylic test piece surface because of formation of nano-sized rugged surfaces. The nanoparticle coating reduces the minimum adhesion force normalized by the particle diameter between the uranium dioxide particle and the acrylic test piece surface, for the smallest particle of about 5 μm associated with desorption, this minimum adhesion force reduced to about 5%. The nanoparticle coating also lowers the adhered plutonium and uranium mixed oxide powder amounts on the acrylic test piece to about 10%. This study reveals that applying the nanoparticle coating to the acrylic panels of the glove box prevents adhesion of nuclear fuel materials. This method effectively reduces the retention of nuclear fuel materials in the glove box, lowers the external exposure dose, and improves the visibility of the acrylic panels.
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9

Kajihara, Katsura, Yasuhiro Aruga, Jun Shimojo, Hiroaki Taniuchi, Tsutomu Takeda, and Masatosi Sasaki. "Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22025.

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New enriched borated aluminum alloys manufactured by melting process are developed, which resulted in supplying structural basket materials for spent nuclear fuel packagings. In this process, the borated aluminum alloys were melted in a vacuum induction furnace at elevated temperature than that of ordinary aluminum melting processes. Boron dissolves into the matrix at the temperature of 1273K or more, and fine aluminum diboride is precipitated and uniformly dispersed upon cooling rapidity. It is confirmed that boron is homogeneously dispersed with the fine particles of approximate 5µm in average size in the product. Tensile strength and creep property at elevated temperature in 1mass-%B 6061-T651 plate and 1mass-%B 3004 extruded rectangular pipe as structural materials are examined. It is confirmed that the both of borated aluminum alloys have stable strength and creep properties that are similar to those of ordinary aluminum alloys.
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Renger, Stefan, Sören Alt, Wolfgang Kästner, André Seeliger, and Frank Zacharias. "Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-54410.

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Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments. This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.
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Reports on the topic "Nuclear fuel surrogate material"

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Connolly, Michael. Aluminum Clad Spent Nuclear Fuel Task 6: Surrogate Sample Preparation and Validation Experiment Test Plan. Office of Scientific and Technical Information (OSTI), September 2018. http://dx.doi.org/10.2172/1469390.

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Toni Y. Gutknecht, Guy L. Fredrickson, and Vivek Utgikar. Thermal Analysis of Surrogate Simulated Molten Salts with Metal Chloride Impurities for Electrorefining Used Nuclear Fuel. Office of Scientific and Technical Information (OSTI), April 2012. http://dx.doi.org/10.2172/1055967.

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Author, Not Given. Material control and accounting in the Department of Energy's nuclear fuel complex. Office of Scientific and Technical Information (OSTI), January 1989. http://dx.doi.org/10.2172/6347772.

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Rudman, K., P. Dickerson, Darrin David Byler, P. Peralta, H. Lim, R. McDonald, R. Dickerson, and Kenneth James Mcclellan. 3D Microstructural Characterization of Uranium Oxide as a Surrogate Nuclear Fuel: Effect of Oxygen Stoichiometry on Grain Boundary Distributions. Office of Scientific and Technical Information (OSTI), September 2017. http://dx.doi.org/10.2172/1392797.

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PACKER, M. J. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 1 Fuel. Office of Scientific and Technical Information (OSTI), November 1999. http://dx.doi.org/10.2172/798703.

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Praga, A. N. 105-K Basin material design basis feed description for spent nuclear fuel project facilities. Office of Scientific and Technical Information (OSTI), January 1998. http://dx.doi.org/10.2172/10148244.

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Wang, Jy-An John, Hao Jiang, and Hong Wang. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties. Office of Scientific and Technical Information (OSTI), July 2014. http://dx.doi.org/10.2172/1148865.

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Cuta, Judith M., Sarah R. Suffield, James A. Fort, and Harold E. Adkins. Thermal performance sensitivity studies in support of material modeling for extended storage of used nuclear fuel. Office of Scientific and Technical Information (OSTI), August 2013. http://dx.doi.org/10.2172/1113602.

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Pearce, K. L. 105-K Basin material design basis feed description for spent nuclear fuel project facilities. Volume 2: Sludge. Office of Scientific and Technical Information (OSTI), August 1998. http://dx.doi.org/10.2172/10148544.

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PEARCE, K. L. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 2 Sludge. Office of Scientific and Technical Information (OSTI), April 2000. http://dx.doi.org/10.2172/803051.

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