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Journal articles on the topic 'Nuclear fuel surrogate material'

Author: Grafiati

Published: 4 June 2021

Last updated: 17 February 2022

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1

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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2

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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3

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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4

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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5

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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6

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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7

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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8

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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9

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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10

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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11

Gausse, Clémence, Calum W. Dunlop, Aidan A. Friskney, Martin C. Stennett, Neil C. Hyatt, and Claire L. Corkhill. "Synthesis, characterisation and preliminary corrosion behaviour assessment of simulant Fukushima nuclear accident fuel debris." MRS Advances 5, no. 1-2 (2020): 65–72. http://dx.doi.org/10.1557/adv.2020.35.

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ABSTRACTNuclear fuel debris generated at the Fukushima Daiichi nuclear power plant during the loss of coolant accident in 2011, still resides within the reactor units, constantly cooled by water. Until it is retrieved, the fuel debris will corrode, releasing radioactive elements into the coolant water and the ground surrounding the reactors. To predict the corrosion behaviour of these materials, and to establish parameters for experiments with U-containing and real fuel debris, the corrosion of two surrogate fuel debris materials, with a composition of Ce(1-x)ZrxO2 (x = 0.2 and 0.4), was investigated. Materials were synthesised by a wet chemistry route and pellets were sintered at 1700°C in air atmosphere. Due to the slow corrosion kinetics, aggressive conditions were applied, and corrosion experiments were performed in 9 mol.L-1 HNO3 under static conditions. The incorporation of Zr into the structure of Ce reduced the normalised dissolution rate; from (3.75 ± 0.15) × 10-6 g.m-2.d-1 to (4.96 ± 0.28) × 10-6 g.m-2.d-1 for RL(Ce) of Ce0.8Zr0.2O2 and Ce0.6Zr0.4O2, respectively.

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12

McDeavitt, S. M., A. Parkison, A. R. Totemeier, and J. J. Wegener. "Fabrication of Cermet Nuclear Fuels Designed for the Transmutation of Transuranic Isotopes." Materials Science Forum 561-565 (October 2007): 1733–36. http://dx.doi.org/10.4028/www.scientific.net/msf.561-565.1733.

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The Uranium Extraction (UREX) family of processes uses solvent extraction techniques designed to partition spent uranium and transuranic (TRU) isotopes from fission product waste. Once separated, the collective TRU elements (Np, Pu, Am, and Cm) can be recycled in advanced nuclear energy systems. A zirconium matrix cermet is proposed as a fuel form for this application. Processing methods have been designed to convert the TRU product and spent Zircaloy cladding into feed materials for the hot extrusion of the cermet fuel pins. The TRU conversion process is being developed using a surrogate mixture of uranium and cerium nitrate solutions to generate mixed oxide microspheres. The Zircaloy recovery process is a hydride-dehydride method that is being demonstrated at the bench scale. The powder products from these methods may be combined through hot extrusion into a cermet composite; demonstration experiments using zirconium powder and zirconia microspheres have been completed.

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13

Hunt, R. D., J. D. Hunn, J. F. Birdwell, T. B. Lindemer, and J. L. Collins. "The addition of silicon carbide to surrogate nuclear fuel kernels made by the internal gelation process." Journal of Nuclear Materials 401, no. 1-3 (June 2010): 55–59. http://dx.doi.org/10.1016/j.jnucmat.2010.03.018.

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14

Pukari, Merja, and Masahide Takano. "Sintering and characterization of ZrN and (Dy,Zr)N as surrogate materials for fast reactor nitride fuel." Journal of Nuclear Materials 444, no. 1-3 (January 2014): 7–13. http://dx.doi.org/10.1016/j.jnucmat.2013.09.001.

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15

Osaka, Masahiko, Shuhei Miwa, and Yoshiaki Tachi. "Simple fabrication process for CeO2–MgO composite as surrogate for actinide-containing target for use in nuclear fuel." Ceramics International 32, no. 6 (January 2006): 659–63. http://dx.doi.org/10.1016/j.ceramint.2005.04.026.

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16

Wang, Hong, and Jy-An John Wang. "Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets." Journal of Nuclear Materials 479 (October 2016): 470–82. http://dx.doi.org/10.1016/j.jnucmat.2016.07.044.

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17

Dewi, Ariyani Kusuma, Souichirou Yamaguchi, Takashi Onitzuka, and Masayoshi Uno. "Thermal conductivity calculation of ZrC–ZrO2 pellet from powder metallurgy as the surrogate of UCO kernel fuel." Journal of Nuclear Materials 539 (October 2020): 152343. http://dx.doi.org/10.1016/j.jnucmat.2020.152343.

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18

Tian, Wei, M. A. Pouchon, Hangxu Guo, Denglei Chen, Xiaojie Yin, and Zhi Qin. "Fabrication of CeO2 ceramic spheres as a surrogate of nuclear fuel by an improved microwave-assisted rapid internal gelation process." Ceramics International 44, no. 6 (April 2018): 6739–46. http://dx.doi.org/10.1016/j.ceramint.2018.01.090.

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19

Kwon, Sang Woon, Si Woo Park, and Sung Jai Lee. "Effect of Deposit on the Evaporation Rate of Adhered Salt in Uranium Dendrite." Science and Technology of Nuclear Installations 2020 (July 8, 2020): 1–6. http://dx.doi.org/10.1155/2020/8866234.

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Electrorefining is a key step in pyroprocessing. The solid cathode processing is necessary to separate the salt from the cathode of the electrorefiner since the uranium deposit in a solid cathode contains electrolyte salt. Moreover, it is very important to increase the throughput of the salt separation system due to the high uranium content of the spent nuclear fuel and high salt fraction of uranium dendrites. Therefore, in this study, the effect of deposit on the evaporation of the adhered salt in a uranium deposit was investigated by using the samples of salt in the uranium deposit and salt in the deposit of the surrogate material for the effective separation of the salt. It was found that the salt evaporation rate is dependent on the deposit type and bulk density in the crucible. Additionally, the evaporation rate was found to be lower when the deposit structure is complex; the rate also decreases as the bulk density of the deposit is increased owing to the retardation of the salt vapour transport process. It was concluded that the mass transfer of the salt vapour is an important parameter for the achievement of a high throughput performance in the salt distillation process.

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20

Kulikova, Svetlana A., Kseniya Yu Belova, Ekaterina A. Tyupina, and Sergey E. Vinokurov. "Conditioning of Spent Electrolyte Surrogate LiCl-KCl-CsCl Using Magnesium Potassium Phosphate Compound." Energies 13, no. 8 (April 16, 2020): 1963. http://dx.doi.org/10.3390/en13081963.

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The current work was aimed at developing a new conditioning method of spent electrolyte-radioactive waste (RW) generated during the pyrochemical reprocessing of mixed nitride uranium-plutonium spent nuclear fuel. Magnesium potassium phosphate (MPP) compound samples were synthesized under solidification of the electrolyte surrogate solution in a LiCl-KCl-CsCl system. The phase composition and structure of obtained compounds were studied by XRD and SEM-EDS methods. It was found that the compounds possessed a high compressive strength of 17–26 MPa. Hydrolytic stability of the compounds was evaluated in accordance with the long semi-dynamic test GOST R 52126-2003 and with the static PCT test. The 137Cs content in the leached solutions was determined by gamma-ray spectrometry, and other compound components were determined by ICP–AES and ICP–MS methods. The differential leaching rate of Cs at 25 °C from monolithic samples on the 91st day of samples contact with water was 5–11 × 10−5 g/(cm2·day) (GOST R 52126-2003), and was 4–29 × 10−7 g/(cm2∙day) on the 7th day at 90 °C from crushed samples (PCT). The thermal stability of the compound at 180 °C and 450 °C was shown. The characteristics of the obtained MPP compound correspond to the current regulatory requirements for materials for RW conditioning.

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21

Prabhu, Sreehari Ramachandra, Mahesh D. Pandey, Nicolas Christodoulou, and Brian W. Leitch. "A surrogate model for the 3D prediction of in-service deformation in CANDU® fuel channels." Nuclear Engineering and Design 369 (December 2020): 110871. http://dx.doi.org/10.1016/j.nucengdes.2020.110871.

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22

Williams, Ammon, Keith Bryce, and Supathorn Phongikaroon. "Measurement of Cerium and Gadolinium in Solid Lithium Chloride–Potassium Chloride Salt Using Laser-Induced Breakdown Spectroscopy (LIBS)." Applied Spectroscopy 71, no. 10 (July 18, 2017): 2302–12. http://dx.doi.org/10.1177/0003702817709298.

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Pyroprocessing of used nuclear fuel (UNF) has many advantages—including that it is proliferation resistant. However, as part of the process, special nuclear materials accumulate in the electrolyte salt and present material accountability and safeguards concerns. The main motivation of this work was to explore a laser-induced breakdown spectroscopy (LIBS) approach as an online monitoring technique to enhance the material accountability of special nuclear materials in pyroprocessing. In this work, a vacuum extraction method was used to draw the molten salt (CeCl3-GdCl3-LiCl-KCl) up into 4 mm diameter Pyrex tubes where it froze. The salt was then removed and the solid salt was measured using LIBS and inductively coupled plasma mass spectroscopy (ICP-MS). A total of 36 samples were made that varied the CeCl3 and GdCl3 (surrogates for uranium and plutonium, respectively) concentrations from 0.5 wt% to 5 wt%. From these samples, univariate calibration curves for Ce and Gd were generated using peak area and peak intensity methods. For Ce, the Ce 551.1 nm line using the peak area provided the best calibration curve with a limit of detection (LOD) of 0.099 wt% and a root mean squared error of cross-validation (RMSECV) of 0.197 wt%. For Gd, the best curve was generated using the peak intensities of the Gd 564.2 nm line resulting in a LOD of 0.027 wt% and a RMSECV of 0.295 wt%. The RMSECV for the univariate cases were determined using leave-one-out cross-validation. In addition to the univariate calibration curves, partial least squares (PLS) regression was done to develop a calibration model. The PLS models yielded similar results with RMSECV (determined using Venetian blind cross-validation with 17% left out per split) values of 0.30 wt% and 0.29 wt% for Ce and Gd, respectively. This work has shown that solid pyroprocessing salt can be qualitatively and quantitatively monitored using LIBS. This work has the potential of significantly enhancing the material monitoring and safeguards of special nuclear materials in pyroprocessing.

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23

Hong, Seong-Gu, Thak-Sang Byun, Lance L. Snead, and Chong Soo Lee. "Evaluation methods for the hoop strength of small-sized tubular ceramic components." Journal of Materials Research 24, no. 4 (April 2009): 1422–34. http://dx.doi.org/10.1557/jmr.2009.0158.

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Miniaturized test methods were developed and applied to measure the hoop strength of small tubular components. A diametrical loading method and an internal pressurization method using elastomeric insert were adopted in the development of testing methods. Detailed analyses to assess these testing methods were attempted by using theoretical solutions and finite element analysis for stress distributions and the characteristics of the Weibull statistics. To demonstrate the applicability of the test methods, commercially available alumina tubes and miniature silicon carbide coatings from surrogate nuclear fuel particles were tested and their fracture strength distributions were analyzed with Weibull statistics. The size scaling relationship on the fracture strength was investigated by correlating with effective surface area. Furthermore, the applicability of the testing methods was discussed in terms of multiaxial stress fields, altered stress distribution by flattened loading contact, and availability of insert, focusing on high-temperature applications.

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24

Jung, Hundal, Tae Ahn, Roberto Pabalan, and David Pickett. "Corrosion Study of SIMFUEL in Aerated Carbonate Solution Containing Calcium and Silicate." MRS Proceedings 1518 (2013): 139–44. http://dx.doi.org/10.1557/opl.2013.77.

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ABSTRACTThe corrosion behavior of simulated spent nuclear fuel (SIMFUEL) was investigated using electrochemical impedance spectroscopy and solution chemistry analyses. The SIMFUEL was exposed to aerated solutions of NaCl+NaHCO3 with and without calcium (Ca) and silicate. Two SIMFUEL compositions were studied, representing spent nuclear fuel (SNF) corresponding to 3 or 6 at % burnup in terms of fission product equivalents of surrogate elements. For all tested cases, the polarization resistance increased with increased immersion time, indicating possible blocking effects due to accumulation of corrosion products on the SIMFUEL surface. The potential-pH diagram suggests formation of schoepite that may cause the increase in the polarization resistance. The addition of Ca and silicate produced no measureable change in the polarization resistance measured at the corrosion potential. The dissolution rate ranged from 1 to 3 mg/m2-day, which is similar to the range of dissolution rates for SIMFUEL and SNF reported in the literature for comparable conditions. SIMFUEL burnup did not have a major effect on the dissolution rate. Analysis of the solution chemistry shows that uranium is the dominant element dissolved in the posttest solutions, and the dissolution rates calculated from uranium (U) concentrations are consistent with the dissolution rates obtained from impedance measurements. Simulated-fission product elements (i.e., barium, molybdenum, strontium, and zirconium) dissolved from the SIMFUEL electrode at a relatively high rate. Sorption test results indicated significant sorption of U onto the oxide formed on stainless steel. Electrochemical methods were found to be effective for measuring the uranium dissolution rate in real time.

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25

Johnson A., Jonathan, Ryan Wilkerson, Stephen DiPietro, and Gregory B. Thompson. "Cermet surrogate nuclear fuels from coated powders." Journal of Nuclear Materials 557 (December 2021): 153246. http://dx.doi.org/10.1016/j.jnucmat.2021.153246.

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Bondarenko, V. N., A. V. Goncharov, I. M. Karnaukhov, A. V. Mazilov, V. M. Pistryak, V. I. Sukhostavets, A. G. Tolstolutskii, and K. G. Rudya. "Investigation of the elemental composition of surrogates of fuel-containing materials by nuclear microanalysis." Atomic Energy 102, no. 4 (April 2007): 304–9. http://dx.doi.org/10.1007/s10512-007-0047-6.

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27

Rosales, Jhonathan, Isabella J. van Rooyen, and Clemente J. Parga. "Characterizing surrogates to develop an additive manufacturing process for U3Si2 nuclear fuel." Journal of Nuclear Materials 518 (May 2019): 117–28. http://dx.doi.org/10.1016/j.jnucmat.2019.02.026.

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28

Nguyen, V. T., S. Wakayama, S. Kishigami, H. Ohta, T. Hikichi, E. Mizutani, H. T. Bui, and T. Wakayama. "139 INJECTION OF SOMATIC CELL CYTOPLASM INTO OOCYTES BEFORE ICSI IMPAIRED FULL-TERM DEVELOPMENT AND INCREASED PLACENTA WEIGHT IN MICE." Reproduction, Fertility and Development 18, no. 2 (2006): 178. http://dx.doi.org/10.1071/rdv18n2ab139.

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During the process of somatic cell nuclear transfer, cytoplasm is introduced into the enucleated oocytes, in addition to the genomic material, regardless of the electrofusion methods (Wilmut et al. 1997) or direct injection of somatic nucleus by the Honolulu method (Wakayama et al. 1998). Only 1 to 2% of cloned embryos, however, develop to term with many incidences of developmental anomalies. These cloning failures may be explained by incomplete reprogramming of the donor cell genome, although it is not yet clear whether cytoplasmic materials of the somatic cell also have an affect on development of the cloned embryo. In an attempt to answer this question, this study investigates the effects of somatic cytoplasm of different mouse strains and cytoplasm of fertilized embryos at different stage by injecting them into intact mouse oocytes before intracytoplasmic sperm injection (ICSI). Mature oocytes collected from B6D2F1 female after 14 to 16 h of hCG injection were injected with (1) B6D2F1 cumulus cell cytoplasm with different volumes (collected by 2 to 3 �m of injection pipette and piezo pulses), (2) cumulus cell cytoplasm from different mouse strains (B6D2F1, ICR, C57BL/6), (3) cytoplasm of 1- to 8-cell embryos. After subsequent culture for 1 h, B6D2F1 sperm were injected into those oocytes and examined for preimplantation developmental competence. The total number of cells, inner cell mass (ICM), and expression of Oct4 in expanded blastocysts were also examined. In order to examine the effects of somatic cytoplasm on full-term development, we transferred 2-cell embryos at 24 h or morula and blastocysts at 72 h after ICSI to the oviduct or uterus of surrogate mothers (ICR) on Day 1 or 3 of pseudopregnancy. The control group received a sham injection with PVP before ICSI. The results showed that an increase the volume of cytoplasm from 1-fold to 4-fold (equivalent with the volume of 1 cumulus cell) resulted in impairing full-term development (28 and 7%, respectively, vs. 56 to 63% in the control group, P < 0.01). There was no difference in the frequency of embryos developing to the blastocyts stage between B6D2F1 and ICR somatic cytoplasms at the same volume. However, C57BL/6 somatic cytoplasm induced the 2-cell block to B6D2F2 embryos. Fertilized embryo cytoplasm did not reduce the frequency of blastocyst stage and full-term development. Interestingly, we found that somatic cytoplasm increased the placenta weight of ICSI embryo (0.2002 � 0.03, n = 32; vs. 0.1198 � 0.02 in control group, n = 87; P < 0.01). We also obtained placenta with no fetus when the volume of somatic cytoplasm was the same size as cumulus cell. We found that an increase in the volume of somatic cytoplasm led to low expression of Oct4 in expanded blastocysts. These findings indicated that injection of somatic cytoplasm into oocytes before ICSI decreased the preimplantation development, clearly impaired full-term development, and caused placental overgrowth in fertilized embryos. This study suggested that somatic cell cytoplasmic material is one cause of the low rate of full-term development of cloned animals.

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29

Worley, Christopher G., and George J. Havrilla. "Micro-X-ray Fluorescence Characterization of Mixed Oxide Fuel Surrogate Feed Material." Analytical Chemistry 70, no. 14 (July 1998): 2957–63. http://dx.doi.org/10.1021/ac9713924.

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30

Latimer, G. D., W. R. Marcum, and W. F. Jones. "Dispersion of Surrogate LWR Fuel Experiments Under LOCA Conditions." Nuclear Technology 206, no. 9 (March 2, 2020): 1374–84. http://dx.doi.org/10.1080/00295450.2020.1712158.

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31

Stefanovsky, Sergey V., Vladimir V. Lebedev, Alexander G. Ptashkin, Sergey A. Dmitriev, and James C. Marra. "Cold Crucible Inductive Melting Technology – Application to Vitrification and Ceramization of High Level and Actinide Wastes." Advances in Science and Technology 73 (October 2010): 183–93. http://dx.doi.org/10.4028/www.scientific.net/ast.73.183.

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Cold crucible inductive melting is a promising method for production of high-temperature materials. The method is based on direct heating of conductive materials by high-frequency (105-107 Hz) electromagnetic field from an external source. Application of the CCIM to production of vitreous borosilicate and alumino/iron phosphate and ceramic waste forms such as Synroc and its varieties and pyrochlore, murataite and garnet-based ceramics has been successfully demonstrated. Currently a full-scale low level waste vitrification plant based on a 418 mm inner diameter cold crucibles energized from a 1.76MHz/160 kW generators is under operation at SIA Radon. This plant was used for demonstration of feasibility of cold crucible vitrification of Savannah River Site high-iron and high iron/aluminum high level wastes. Numerous ceramic and glass ceramic materials containing high level and actinide waste surrogates such as actinide and actinide/rare earth fractions of high level waste have been successfully produced in the Radon lab- and bench-scale cold crucible based units operated at 5.28 and 1.76 MHz. Large-scale cold crucibles may be applied for vitrification of liquid and solid low and high level wastes whereas small-scale cold crucible may be efficiently used for immobilization of actinide-bearing waste generated from advanced nuclear fuel cycle reprocessing.

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32

Fantidis, J. G., G. E. Nikolaou, and F. N. Tsagas. "Identification of unknown nuclear material." HNPS Proceedings 15 (January 1, 2020): 273. http://dx.doi.org/10.12681/hnps.2611.

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The identification of spent PWR nuclear fuel in terms of its initial enrichment and final burnup is demonstrated. Spent UO2 fuel from a PWR power station was used as the nuclear material of supposed unknown irradiation history. The identification procedure was based on determining the U and Pu isotopic composition of the fuel by chemical analyses, simulation calculations of fuel evolution and statistical analysis. The procedures followed and associated limitations are discussed.

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Figueroa, Antonio, and Malte Göttsche. "Gaussian processes for surrogate modeling of discharged fuel nuclide compositions." Annals of Nuclear Energy 156 (June 2021): 108085. http://dx.doi.org/10.1016/j.anucene.2020.108085.

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Emond, Christy A., and John F. Kalinich. "Biokinetics of Embedded Surrogate Radiological Dispersal Device Material." Health Physics 102, no. 2 (February 2012): 124–36. http://dx.doi.org/10.1097/hp.0b013e31823095e5.

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Patnaik, Sobhan. "Comparative analysis of temperature dependent properties of commercial nuclear fuel pellet and surrogates undergoing cracking: A review." Ceramics International 46, no. 16 (November 2020): 24765–78. http://dx.doi.org/10.1016/j.ceramint.2020.06.266.

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36

Terricabras, Adrien J., James O. Kiggans, Ling Wang, and Steven J. Zinkle. "Characterization of high thermal conductivity fuel surrogates before and after ion irradiation." Journal of Nuclear Materials 552 (August 2021): 153027. http://dx.doi.org/10.1016/j.jnucmat.2021.153027.

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37

Zhang, Ling Zhe, Ya Kun Sun, Su Li, and Qing Ping Zheng. "Simulation of HCCI Combustion Characteristics for Low RON Gasoline Surrogate Fuels." Applied Mechanics and Materials 694 (November 2014): 54–58. http://dx.doi.org/10.4028/www.scientific.net/amm.694.54.

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A reduced chemical kinetic model (103species and 468 reactions) for new low-RON(research octane number) gasoline surrogate fuels has been proposed. Simulations explored for ignition delay time have been compared with experimental data in shock tubes at pressure of 10atm-55 atm and temperatue of 600-1400 K (fuel/air equivalence ratio=0.5,1.0,2.0 and EGR rate=0, 20%). The simulation data presented 15% enlargement compared with experiments showed applicability of the new kinetic mode in this work. A combustion simulation model has been build for HCCI(homogeneous charge compression ignition) engine with Chemkin-pro. The effects of different air inlet temperature, inlet pressure, engine speed and the fuel air equivalence ratio on the combustion characteristics of the fuel were researched. The results indicated the combustion in an HCCI engine worked sufficiently with lean mixtures and low speed. Meanwhile the material strength could be influenced when the inlet conditions changed. This helps to promote the low-RON gasoline surrogate fuel application in the HCCI engine.

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AKAOKA, Katsuaki, Masabumi MIYABE, Haruyoshi OTOBE, and Ikuo WAKAIDA. "Laser-Induced Breakdown Spectroscopy for Nuclear Fuel Material." Review of Laser Engineering 42, no. 12 (2014): 918. http://dx.doi.org/10.2184/lsj.42.12_918.

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39

Goltsev, V. Y., and A. V. Osintsev. "Estimating the Brittle Strength of Nuclear Fuel Material." KnE Materials Science 4, no. 1 (May 6, 2018): 139. http://dx.doi.org/10.18502/kms.v4i1.2138.

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40

Lützenkirchen, Klaus, Maria Wallenius, Zsolt Varga, Thierry Wiss, Alexander Knott, Adrian Nicholl, and Klaus Mayer. "Nuclear forensics on uranium fuel pellets." Radiochimica Acta 107, no. 7 (July 26, 2019): 635–43. http://dx.doi.org/10.1515/ract-2018-3068.

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Abstract Nuclear forensic analysis of uranium oxide fuel pellets and the respective data interpretation have proven to be important elements of the investigative process when such material is found out of regulatory control. Uranium oxide fuel pellets are produced at industrial scale and have to meet tight technical specifications. Variations in the production process of different manufacturers or fuel fabrication facilities offer an additional source of information. The compilation of such information in a nuclear materials database has proven to be a highly valuable resource. In consequence, measureable material properties (referred to as “signatures”) such as the isotopic composition, the physical dimensions, the chemical impurities, the products of the radioactive decay of uranium and the microstructure allow establishing the processing history of the material. To this end, dedicated analytical methods have to be developed and suitable methods for data evaluation and interpretation need to be utilised. The successful application of these nuclear forensic tools is illustrated with two case studies on real incidents of illicit nuclear trafficking of uranium oxide fuel pellets.

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Azevedo, C. R. F. "Selection of fuel cladding material for nuclear fission reactors." Engineering Failure Analysis 18, no. 8 (December 2011): 1943–62. http://dx.doi.org/10.1016/j.engfailanal.2011.06.010.

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42

Hoggan, Rita E., Larry D. Zuck, W. Roger Cannon, and Paul A. Lessing. "Processing of surrogate nuclear fuel pellets for better dimensional control with dry bag isostatic pressing." Journal of Nuclear Materials 482 (December 2016): 34–41. http://dx.doi.org/10.1016/j.jnucmat.2016.05.034.

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YAMAWAKI, Michio, Hiroo NAKAMURA, Nobuaki NODA, Kenji OKUNO, Kenji NODA, and Satoru TANAKA. "Topics of Fusion Fuel-Material Interactions (FFMI)." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 36, no. 4 (1994): 301–10. http://dx.doi.org/10.3327/jaesj.36.301.

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44

Wang, F., ZL Zheng, and ZW He. "A Soot Precursor Formation Embedded Reaction Mechanism of Diesel Surrogate Fuel." Energy Sources, Part A: Recovery, Utilization, and Environmental Effects 37, no. 12 (May 7, 2015): 1323–31. http://dx.doi.org/10.1080/15567036.2011.610867.

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45

Lee, Young-Woo, Chang Young Joung, Si Hyung Kim, and Sang-Chul Lee. "Inert matrix fuel — A new challenge for material technology in the nuclear fuel cycle." Metals and Materials International 7, no. 2 (April 2001): 159–64. http://dx.doi.org/10.1007/bf03026954.

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46

Le Roux, S. D., and D. J. Van der Merwe. "Texture Analysis in Zircaloy Cladding Tube Material for Nuclear Fuel." Materials Science Forum 157-162 (May 1994): 1455–62. http://dx.doi.org/10.4028/www.scientific.net/msf.157-162.1455.

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47

YONEKAWA, Hidemi, and Kenji OMURA. "Material handling and automation technology at nuclear fuel fabrication plant." Journal of the Japan Society for Precision Engineering 57, no. 5 (1991): 803–6. http://dx.doi.org/10.2493/jjspe.57.803.

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48

Kim, Hyun-Jung, Jeong-Sik Yim, Yong-Jin Jeong, and Kang-Hee Lee. "A study on the mechanically equivalent surrogate plate of U Mo dispersion fuel using tungsten." Nuclear Engineering and Technology 51, no. 2 (April 2019): 495–500. http://dx.doi.org/10.1016/j.net.2018.10.015.

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49

Buck, Edgar C., Nancy L. Dietz, and John K. Bates. "Corroded spent nuclear fuel examined with EELS." Proceedings, annual meeting, Electron Microscopy Society of America 54 (August 11, 1996): 562–63. http://dx.doi.org/10.1017/s0424820100165276.

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Direct disposal of spent nuclear fuel (SNF) into the proposed unsaturated geologic repository at Yucca Mountain, NV is being studied at several laboratories, including Argonne National Laboratory. Corrosion tests with SNF are being conducted to understand the long-term behavior of SNF under conditions designed to simulate the unsaturated conditions at the site. The SNF used in this study was the Approved Testing Material (ATM)-106 with a bum-up of 43 MW·d/kg U. A sample of ATM-106 fuel was exposed to dripping simulated groundwater for 271 days; after this time the experiment was terminated and the material removed for further study. Details of the testing methodology have been given by Finn et al.,.Previous attempts to study SNF with TEM have used ion milled samples, in this study we prepared the samples by ultramicrotomy which reduced the radiological hazard substantially.

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50

Ebiwonjumi, Bamidele, Peng Zhang, and Deokjung Lee. "SENSITIVITY ANALYSIS OF PWR SPENT FUEL DUE TO MODELLING PARAMETER UNCERTAINTIES USING SURROGATE MODELS." EPJ Web of Conferences 247 (2021): 15009. http://dx.doi.org/10.1051/epjconf/202124715009.

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In the BEPU (Best Estimate Plus Uncertainty) framework, uncertainty quantification (UQ) is a requirement to improve confidence and reliability of code predictions. Over the years, a lot of works have been done to quantify uncertainties in code predictions of spent nuclear fuel (SNF) characteristics due to nuclear data uncertainties. The purpose of this study is to quantify the uncertainty in pressurized water reactor (PWR) fuel assembly radiation source terms (isotopic inventory, activity, decay heat, neutron and gamma source) due to uncertainties in modeling parameters. The deterministic code STREAM is used to predict the source terms of a typical PWR fuel assembly following realistic and detailed irradiation history. For the sensitivity analysis (SA) and UQ, surrogate models are developed based on polynomial chaos expansion (PCE) and variance-based global sensitivity indices (i.e., Sobol’ indices) are employed. The global SA identifies the less important uncertain parameters, showing that the number of uncertain input parameters can be reduced. The surrogate model offers a significantly reduced computational burden even with large number of samples required for the SA/UQ of the model response.

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