Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

On the combustion mechanisms of ZrH2 in double-base propellant.

PubMed

Yang, Yanjing; Zhao, Fengqi; Yuan, Zhifeng; Wang, Ying; An, Ting; Chen, Xueli; Xuan, Chunlei; Zhang, Jiankan

2017-12-13

Metal hydrides are regarded as a series of promising hydrogen-supplying fuel for solid rocket propellants. Their effects on the energetic and combustion performances of propellants are closely related to their reaction mechanisms. Here we report a first attempt to determine the reaction mechanism of ZrH 2 , a high-density metal hydride, in the combustion of a double-base propellant to evaluate its potential as a fuel. ZrH 2 is determined to possess good resistance to oxidation by nitrocellulose and nitroglycerine. Thus its combustion starts with dehydrogenation to generate H 2 and metallic Zr. Subsequently, the newly formed Zr and H 2 participate in the combustion and, especially, Zr melts and then combusts on the burning surface which favors the heat feedback to the propellant. This phenomenon is completely different from the combustion behavior of the traditional fuel Al, where the Al particles are ejected off the burning surface of the propellant to get into the luminous flame zone to burn. The findings in this work validate the potential of ZrH 2 as a hydrogen-supplying fuel for double-base propellants.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

Transformation behavior of the γU(Zr,Nb) phase under continuous cooling conditions

NASA Astrophysics Data System (ADS)

Komar Varela, C. L.; Gribaudo, L. M.; González, R. O.; Aricó, S. F.

2014-10-01

The selected alloy for designing a high-density monolithic-type nuclear fuel with U-Zr-Nb alloy as meat and Zry-4 as cladding, has to remain in the γU(Zr,Nb) phase during the whole fabrication process. Therefore, it is necessary to define a range of concentrations in which the γU(Zr,Nb) phase does not decompose under the process conditions. In this work, several U alloys with concentrations between 28.2-66.9 at.% Zr and 0-13.3 at.% Nb were fabricated to study the possible transformations of the γU(Zr,Nb) phase under different continuous cooling conditions. The results of the electrical resistivity vs temperature experiments are presented. For a cooling rate of 4 °C/min a linear regression was determined by fitting the starting decomposition temperature as a function of Nb concentration. Under these conditions, a concentration of 45.3 at.% Nb would be enough to avoid any transformation of the γU(Zr,Nb) phase. In experiments that involve higher cooling conditions, it has been determined that this concentration can be halved.

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

DOE PAGES

Yan, Y.; Qian, S.; Littrell, K.; ...

2015-02-13

A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distributionmore » of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor will be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.« less

Characterization of Ceramic Plasma-Sprayed Coatings, and Interaction Studies Between U-Zr Fuel and Ceramic Coated Interface at an Elevated Temperature

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ki Hwan Kim; Chong Tak Lee; R. S. Fielding

2011-08-01

Candidate coating materials for re-usable metallic nuclear fuel crucibles, HfN, TiC, ZrC, and Y2O3, were plasma-sprayed onto niobium substrates. The coating microstructure and the thermal cycling behavior were characterized, and U-Zr melt interaction studies carried out. The Y2O3 coating layer had a uniform thickness and was well consolidated with a few small pores scattered throughout. While the HfN coating was not well consolidated with a considerable amount of porosity, but showed somewhat uniform thickness. Thermal cycling tests on the HfN, TiC, ZrC, and Y2O3 coatings showed good cycling characteristics with no interconnected cracks forming even after 20 cycles. Interaction studiesmore » done on the coated samples by dipping into a U-20wt.%Zr melt indicated that HfN and Y2O3 did not form significant reaction layers between the melt and the coating while the TiC and the ZrC coatings were significantly degraded. Y2O3 exhibited the most promising performance among HfN, TiC, ZrC, and Y2O3 coatings.« less

Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.; Keiser, Dennis D.

2015-05-01

A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

Structural, microstructural and thermal analysis of U-(6-x)Zr-xNb alloys (x = 0, 2, 4, 6)

NASA Astrophysics Data System (ADS)

Kaity, Santu; Banerjee, Joydipta; Parida, S. C.; Bhasin, Vivek

2018-06-01

Uranium-rich U-Zr-Nb alloy is considered as a good alternative fuel for fast reactors from the perspective of excellent dimensional stability and desired thermo-physical properties to achieve higher burnup. Detailed investigations related to the structural and microstructural characterization, thermal expansion, phase transformation, microhardness were carried out on U-6Zr, U-4Zr-2Nb, U-2Zr-4Nb and U-6Nb alloys (composition in wt%) where the total amount of alloying elements was restricted to 6 wt%. Structural, microstructural and thermal analysis studies revealed that these alloys undergo a series of transformations from high temperature bcc γ-phase to a variety of equilibrium and intermediate phases depending upon alloy composition, cooling rate and quenching. The structural analysis was carried out by Rietveld refinement. The data of U-Nb and U-Zr-Nb alloys have been highlighted and compared with binary U-Zr alloy.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Advanced chemical hydride-based hydrogen generation/storage system for fuel cell vehicles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breault, R.W.; Rolfe, J.

1998-08-01

Because of the inherent advantages of high efficiency, environmental acceptability, and high modularity, fuel cells are potentially attractive power supplies. Worldwide concerns over clean environments have revitalized research efforts on developing fuel cell vehicles (FCV). As a result of intensive research efforts, most of the subsystem technology for FCV`s are currently well established. These include: high power density PEM fuel cells, control systems, thermal management technology, and secondary power sources for hybrid operation. For mobile applications, however, supply of hydrogen or fuel for fuel cell operation poses a significant logistic problem. To supply high purity hydrogen for FCV operation, Thermomore » Power`s Advanced Technology Group is developing an advanced hydrogen storage technology. In this approach, a metal hydride/organic slurry is used as the hydrogen carrier and storage media. At the point of use, high purity hydrogen will be produced by reacting the metal hydride/organic slurry with water. In addition, Thermo Power has conceived the paths for recovery and regeneration of the spent hydride (practically metal hydroxide). The fluid-like nature of the spent hydride/organic slurry will provide a unique opportunity for pumping, transporting, and storing these materials. The final product of the program will be a user-friendly and relatively high energy storage density hydrogen supply system for fuel cell operation. In addition, the spent hydride can relatively easily be collected at the pumping station and regenerated utilizing renewable sources, such as biomass, natural, or coal, at the central processing plants. Therefore, the entire process will be economically favorable and environmentally friendly.« less

Interdiffusion and reaction between U and Zr

NASA Astrophysics Data System (ADS)

Park, Y.; Newell, R.; Mehta, A.; Keiser, D. D.; Sohn, Y. H.

2018-04-01

The microstructural development and diffusion kinetics were examined for the binary U vs. Zr system using solid-to-solid diffusion couples, U vs. Zr, annealed at 580 °C for 960 h, 650 °C for 480 h, 680 °C for 240 h, and 710 °C for 96 h. Scanning and transmission electron microscopies with X-ray energy dispersive spectroscopy were employed for detailed microstructural and compositional analyses. Interdiffusion and reaction in U vs. Zr diffusion couples primarily produced: δ-UZr2 solid solution (hP3) and α‧-U at 580 °C; and (γU,βZr) solid solution (cI2) and α‧-U at 650°, 680° and 710 °C. The α‧-phase was confirmed as a reduced variant of the α-U orthorhombic structure with lattice parameters, a × b × c = 2.65 × 5.40 × 4.75 (Å) with a negligible solubility for Zr at room temperature. Concentration profiles were examined to determine interdiffusion coefficients, integrated interdiffusion coefficients, and intrinsic diffusion coefficients using Boltzmann-Matano, Wagner, and Heumann analyses, respectively. Composition-dependence of interdiffusion coefficients were documented for α-U, δ-UZr2 (at 580 °C) and (γU,βZr) solid solution (at 650°, 680° and 710 °C). U was determined to intrinsically diffuse faster than Zr, approximately by an order of magnitude, in the δ-UZr2 at 580 °C, and (γU,βZr) phases at 650°, 680° and 710 °C. Based on Darken's approach, thermodynamic data available in literature were coupled to estimate the tracer diffusion coefficients and atomic mobilities of U and Zr.

METHOD OF FABRICATING A URANIUM-ZIRCONIUM HYDRIDE REACTOR CORE

DOEpatents

Weeks, I.F.; Goeddel, W.V.

1960-03-22

A method is described of evenly dispersing uranlum metal in a zirconium hydride moderator to produce a fuel element for nuclear reactors. According to the invention enriched uranium hydride and zirconium hydride powders of 200 mesh particle size are thoroughly admixed to form a mixture containing 0.1 to 3% by weight of U/sup 235/ hydride. The mixed powders are placed in a die and pressed at 100 tons per square inch at room temperature. The resultant compacts are heated in a vacuum to 300 deg C, whereby the uranium hydride deoomposes into uranium metal and hydrogen gas. The escaping hydrogen gas forms a porous matrix of zirconium hydride, with uramum metal evenly dispersed therethrough. The advantage of the invention is that the porosity and uranium distribution of the final fuel element can be more closely determined and controlled than was possible using prior methods of producing such fuel ele- ments.

Dimensionally stable metallic hydride composition

DOEpatents

Heung, Leung K.

1994-01-01

A stable, metallic hydride composition and a process for making such a composition. The composition comprises a uniformly blended mixture of a metal hydride, kieselguhr, and a ballast metal, all in the form of particles. The composition is made by subjecting a metal hydride to one or more hydrogen absorption/desorption cycles to disintegrate the hydride particles to less than approximately 100 microns in size. The particles are partly oxidized, then blended with the ballast metal and the kieselguhr to form a uniform mixture. The mixture is compressed into pellets and calcined. Preferably, the mixture includes approximately 10 vol. % or more kieselguhr and approximately 50 vol. % or more ballast. Metal hydrides that can be used in the composition include Zr, Ti, V, Nb, Pd, as well as binary, tertiary, and more complex alloys of La, Al, Cu, Ti, Co, Ni, Fe, Zr, Mg, Ca, Mn, and mixtures and other combinations thereof. Ballast metals include Al, Cu and Ni.

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan

2013-07-01

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order tomore » prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.« less

Contribution to the thermodynamic description of the corium - The U-Zr-O system

NASA Astrophysics Data System (ADS)

Quaini, A.; Guéneau, C.; Gossé, S.; Dupin, N.; Sundman, B.; Brackx, E.; Domenger, R.; Kurata, M.; Hodaj, F.

2018-04-01

In order to understand the stratification process that may occur in the late phase of the fuel degradation during a severe accident in a PWR, the thermodynamic knowledge of the U-Zr-O system is crucial. The presence of a miscibility gap in the U-Zr-O liquid phase may lead to a stratified configuration, which will impact the accidental scenario management. The aim of this work was to obtain new experimental data in the U-Zr-O liquid miscibility gap. New tie-line data were provided at 2567 ± 25 K. The related thermodynamic models were reassessed using present data and literature values. The reassessed model will be implemented in the TAF-ID international database. The composition and density of phases potentially formed during stratification will be predicted by coupling current thermodynamic model with thermal-hydraulics codes.

Studies of hydride formation and superconductivity in hydrides of alloys Th-M /M = La, Y, Ce, Zr and Bi/

NASA Technical Reports Server (NTRS)

Oesterreicher, H.; Clinton, J.; Misroch, M.

1977-01-01

In order to gain a better insight into both the unusual composition of ThH15 and its superconductivity, an experimental study was conducted to assess the influence of partial replacement of Th in Th4H15 by elements which allow for a systematic alteration of spatial and electronic effects. For this purpose, substituent elements with the same number of valence electrons (4) but of smaller size (Zr) as well as elements with a smaller number of valence electrons (3) and either larger (La) or smaller size (Y) were selected. A few data with Ce and Bi as substituent atoms are also included. The matrix alloys for hydriding were obtained by induction melting under Ar in water-cooled Cu boats. Superconducting transition temperatures are found to decrease on substitution for Th in Th4H15. Hydrides derived from LaH3 by substitution for La by Th do not become superconducting. It is suggested that superconductivity in Th4H15 is connected with a deviation from the exact stoichiometry of Th4H15. A model of unsatisfied valencies may be of more general validity in predicting superconductivity.

Studies of behavior of the fuel compound based on the U-Zr micro-heterogeneous quasialloy during cyclic thermal tests

NASA Astrophysics Data System (ADS)

Zaytsev, D. A.; Repnikov, V. M.; Soldatkin, D. M.; Solntsev, V. A.

2017-11-01

This paper provides the description of temperature cycle testing of U-Zr heterogeneous fuel composition. The composition is essentially a niobium-doped zirconium matrix with metallic uranium filaments evenly distributed over the cross section. The test samples 150 mm long had been fabricated using a fiber-filament technology. The samples were essentially two-bladed spiral mandrel fuel elements parts. In the course of experiments the following temperatures were applied: 350, 675, 780 and 1140 °C with total exposure periods equal to 200, 30, 30 and 6 hours respectively. The fuel element samples underwent post-exposure material science examination including: geometry measurements, metallographic analysis, X-ray phase analysis and electron-microscopic analysis as well as micro-hardness measurement. It has been found that no significant thermal swelling of the samples occurs throughout the whole temperature range from 350 °C up to 1140 °C. The paper presents the structural changes and redistribution of the fuel component over the fuel element cross section with rising temperature.

Hydridable material for the negative electrode in a nickel-metal hydride storage battery

DOEpatents

Knosp, Bernard; Bouet, Jacques; Jordy, Christian; Mimoun, Michel; Gicquel, Daniel

1997-01-01

A monophase hydridable material for the negative electrode of a nickel-metal hydride storage battery with a "Lave's phase" structure of hexagonal C14 type (MgZn.sub.2) has the general formula: Zr.sub.1-x Ti.sub.x Ni.sub.a Mn.sub.b Al.sub.c Co.sub.d V.sub.e where ##EQU1##

Influence of hydride orientation on fracture toughness of CWSR Zr-2.5%Nb pressure tube material between RT and 300 °C

NASA Astrophysics Data System (ADS)

Sharma, Rishi K.; Sunil, Saurav; Kumawat, B. K.; Singh, R. N.; Tewari, Asim; Kashyap, B. P.

2017-05-01

An experimental setup was designed, fabricated and used to form radial hydrides in Zr-2.5%Nb alloy pressure tube spool. The design of setup was based on ensuring a hoop stress in the spool greater than threshold stress for reorientation of hydrides in this alloy, which was achieved by manipulating the thermal expansion coefficient of the plunger and pressure tube material and diametral interference between them. The experimental setup was loaded on a universal testing machine (UTM) fitted with an environmental chamber and subjected to a temperature cycle for the stress reorientation treatment. The metallographic examination of the hydrogen charged spools subjected to stress re-orientation treatment using this set up revealed formation of predominantly radial hydrides. The variation of fracture toughness of material containing radial hydride with test temperature showed typical 'S' curve behavior with transition temperatures more than that of the material containing circumferential hydride.

Thermophysicochemical Reaction of ZrCo-Hydrogen-Helium System

NASA Astrophysics Data System (ADS)

Jung, Kwangjin; Kang, Hee-Seok; Yun, Sei-Hun; Chung, Hongsuk

2017-11-01

Nuclear fusion energy, which is clean and infinite, has been studied for more than half a century. Efforts are in progress worldwide for the demonstration and validation of nuclear fusion energy. Korea has been developing hydrogen isotope storage and delivery system (SDS) technologies including a basic scientific study on a hydrogen storage medium. An SDS bed, which is a key component of the SDS, is used for storing hydrogen isotopes in a metal hydride form and supplying them to a tokamak. Thermophysicochemical properties of the ZrCo-H2-He system are investigated for the practical utilization of a hydriding alloy system. The hydriding reaction, in which ZrCoHx is composed as ZrCo absorbing hydrogen, is exothermic. The dehydriding reaction, in which ZrCoHx decomposes into ZrCo and hydrogen, is endothermic. The heat generated through the hydriding reaction interrupts the hydriding progress. The heat loss by a dehydriding reaction impedes the dehydriding progress. The tritium decay product, helium-3, covers the ZrCo and keeps the hydrogen from contact with ZrCo in the SDS bed. In this study, we designed and fabricated a ZrCo bed and its performance test rig. The helium blanketing effect on a ZrCo hydrogen reaction with 0 % to 20 % helium content in a gaseous phase and a helium blanket removal method were studied experimentally. In addition, the volumetric flow rates and temperature at the beginning of a ZrCo hydrogen reaction in a hydrogen or helium atmosphere, and the cooling of the SDS bed by radiation only and by both radiation and natural convection related to the reuse cycle, were obtained.

Identification of phases in the interaction layer between U-Mo-Zr/Al and U-Mo-Zr/Al-Si

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varela, C.L. Komar; Arico, S.F.; Mirandou, M.

Out-of-pile diffusion experiments were performed between U-7wt.% Mo-1wt.% Zr and Al or Al A356 (7,1wt.% Si) at 550 deg. C. In this work morphological characterization and phase identification on both interaction layer are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-Ray diffraction and WDS microanalysis. In the interaction layer U-7wt.% Mo-1wt.% Zr/Al, the phases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U and Al{sub 43}Mo{sub 4}U{sub 6} were identified. In the interaction layer U-7wt.% Mo-1wt.% Zr/Al A356, the phases U(Al, Si) with 25at.% Si and Si{sub 5}U{sub 3} were identified. This lastmore » phase, with a higher Si concentration, was identified with XRD Synchrotron radiation performed at the National Synchrotron Light Laboratory (LNLS), Campinas, Brasil. (author)« less

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

NASA Astrophysics Data System (ADS)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Nanoindentation study of bulk zirconium hydrides at elevated temperatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinbiz, Mahmut Nedim; Balooch, Mehdi; Hu, Xunxiang

Here, the mechanical properties of zirconium hydrides was studied using nano-indentation technique at a temperature range of 25 – 400 °C. Temperature dependency of reduced elastic modulus and hardness of δ- and ε-zirconium hydrides were obtained by conducting nanoindentation experiments on the bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of δ-zirconium hydride (H/Zr ratio =1.61) decreased from ~113 GPa to ~109 GPa while temperature increased from room temperature to 400°C. For ε-zirconium hydrides (H/Zr ratio=1.79), the reduced elastic modulus decreased from 61 GPa to 54 GPa as temperature increased from roommore » temperature to 300 °C. Whereas, hardness of δ-zirconium hydride significantly decreased from 4.1 GPa to 2.41 GPa when temperature increased from room temperature to 400 °C. Similarly, hardness of ε-zirconium hydride decreased from 3.06 GPa to 2.19 GPa with temperature increase from room temperature to 300°C.« less

Nanoindentation study of bulk zirconium hydrides at elevated temperatures

DOE PAGES

Cinbiz, Mahmut Nedim; Balooch, Mehdi; Hu, Xunxiang; ...

2017-08-02

Here, the mechanical properties of zirconium hydrides was studied using nano-indentation technique at a temperature range of 25 – 400 °C. Temperature dependency of reduced elastic modulus and hardness of δ- and ε-zirconium hydrides were obtained by conducting nanoindentation experiments on the bulk hydride samples with independently heating capability of indenter and heating stage. The reduced elastic modulus of δ-zirconium hydride (H/Zr ratio =1.61) decreased from ~113 GPa to ~109 GPa while temperature increased from room temperature to 400°C. For ε-zirconium hydrides (H/Zr ratio=1.79), the reduced elastic modulus decreased from 61 GPa to 54 GPa as temperature increased from roommore » temperature to 300 °C. Whereas, hardness of δ-zirconium hydride significantly decreased from 4.1 GPa to 2.41 GPa when temperature increased from room temperature to 400 °C. Similarly, hardness of ε-zirconium hydride decreased from 3.06 GPa to 2.19 GPa with temperature increase from room temperature to 300°C.« less

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant

Mechanical properties of zirconium alloys and zirconium hydrides predicted from density functional perturbation theory

DOE PAGES

Weck, Philippe F.; Kim, Eunja; Tikare, Veena; ...

2015-10-13

Here, the elastic properties and mechanical stability of zirconium alloys and zirconium hydrides have been investigated within the framework of density functional perturbation theory. Results show that the lowest-energy cubic Pn-3m with combining macron]m polymorph of δ-ZrH 1.5 does not satisfy all the Born requirements for mechanical stability, unlike its nearly degenerate tetragonal P4 2/ mcm polymorph. Elastic moduli predicted with the Voigt–Reuss–Hill approximations suggest that mechanical stability of α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates is limited by the shear modulus. According to both Pugh's and Poisson's ratios, α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates can be considered ductile. The Debyemore » temperatures predicted for γ-ZrH, δ-ZrH 1.5 and ε-ZrH 2 are θ D = 299.7, 415.6 and 356.9 K, respectively, while θ D = 273.6, 284.2, 264.1 and 257.1 K for the α-Zr, Zry-4, ZIRLO and M5 matrices, i.e. suggesting that Zry-4 possesses the highest micro-hardness among Zr matrices.« less

SEM in situ MiniCantilever Beam Bending of U-10Mo/Zr/Al Fuel Elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mook, William; Baldwin, Jon K.; Martinez, Ricardo M.

2014-06-16

In this work, the fracture behavior of Al/Zr and Zr/dU-10Mo interfaces was measured via the minicantilever bend technique. The energy dissipation rates were found to be approximately 3.7-5 mj/mm 2 and 5.9 mj/mm 2 for each interface, respectively. It was found that in order to test the Zr/U-10Mo interface, location of the hinge of the cantilever was a key parameter. While this test could be adapted to hot cell use through careful alignment fixturing and measurement of crack lengths with an optical microscope (as opposed to SEM, which was used here out of convenience), machining of the cantilevers via MiniMillmore » in such a way as to locate the interfaces at the cantilever hinge, as well as proper placement of a femtosecond laser notch will continue to be key challenges in a hot cell environment.« less

Determination of 90Sr / 238U ratio by double isotope dilution inductively coupled plasma mass spectrometer with multiple collection in spent nuclear fuel samples with in situ 90Sr / 90Zr separation in a collision-reaction cell

NASA Astrophysics Data System (ADS)

Isnard, H.; Aubert, M.; Blanchet, P.; Brennetot, R.; Chartier, F.; Geertsen, V.; Manuguerra, F.

2006-02-01

Strontium-90 is one of the most important fission products generated in nuclear industry. In the research field concerning nuclear waste disposal in deep geological environment, it is necessary to quantify accurately and precisely its concentration (or the 90Sr / 238U atomic ratio) in irradiated fuels. To obtain accurate analysis of radioactive 90Sr, mass spectrometry associated with isotope dilution is the most appropriated method. But, in nuclear fuel samples the interference with 90Zr must be previously eliminated. An inductively coupled plasma mass spectrometer with multiple collection, equipped with an hexapole collision cell, has been used to eliminate the 90Sr / 90Zr interference by addition of oxygen in the collision cell as a reactant gas. Zr + ions are converted into ZrO +, whereas Sr + ions are not reactive. A mixed solution, prepared from a solution of enriched 84Sr and a solution of enriched 235U was then used to quantify the 90Sr / 238U ratio in spent fuel sample solutions using the double isotope dilution method. This paper shows the results, the reproducibility and the uncertainties that can be obtained with this method to quantify the 90Sr / 238U atomic ratio in an UOX (uranium oxide) and a MOX (mixed oxide) spent fuel samples using the collision cell of an inductively coupled plasma mass spectrometer with multiple collection to perform the 90Sr / 90Zr separation. A comparison with the results obtained by inductively coupled plasma mass spectrometer with multiple collection after a chemical separation of strontium from zirconium using a Sr spec resin (Eichrom) has been performed. Finally, to validate the analytical procedure developed, measurements of the same samples have been performed by thermal ionization mass spectrometry, used as an independent technique, after chemical separation of Sr.

Calculations of hydrogen diffusivity in Zr-based alloys: Influence of alloying elements and effect of stress

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yu, J.; Jiang, C.; Zhang, Y.

This report summarizes the progress on modeling hydrogen diffusivity in Zr-based alloys. The presence of hydrogen (H) can detrimentally affect the mechanical properties of many metals and alloys. To mitigate these detrimental effects requires fundamental understanding of the thermodynamics and kinetics governing H pickup and hydride formation. In this work, we focus on H diffusion in Zr-based alloys by studying the effects of alloying elements and stress, factors that have been shown to strongly affect H pickup and hydride formation in nuclear fuel claddings. A recently developed accelerated kinetic Monte Carlo method is used for the study. It is foundmore » that for the alloys considered here, H diffusivity depends weakly on composition, with negligible effect at high temperatures in the range of 600-1200 K. Therefore, the small variation in compositions of these alloys is likely not a major cause of the very different H pickup rates. In contrast, stress strongly affects H diffusivity. This effect needs to be considered for studying hydride formation and delayed hydride cracking.« less

Oxidation/reduction studies on Zr yU 1-yO 2+x and delineation of a new orthorhombic phase in U-Zr-O system

NASA Astrophysics Data System (ADS)

Sali, S. K.; Kulkarni, N. K.; Krishnan, K.; Achary, S. N.; Tyagi, A. K.

2008-08-01

In this communication, we report the oxidation and reduction behavior of fluorite type solid solutions in U-Zr-O. The maximum solubility of ZrO 2 in UO 2 lattice could be achieved with a mild oxidizing followed by reducing conditions. The role of valency state of U is more dominating in controlling the unit cell parameters than the incorporated interstitial oxygen in the fluorite lattice. The controlled oxidation studies on U-Zr-O solid solutions led to the delineation of a new distorted fluorite lattice at the U:Zr=2:1 composition. The detailed crystal structure analysis of this ordered composition Zr 0.33U 0.67O 2.33 (ZrU 2O 7) has been carried from the powder XRD data. This phase crystallizes in an orthorhombically distorted fluorite type lattice with unit cell parameters: a=5.1678(2), b=5.4848(2), c=5.5557(2) Å and V=157.47(1) Å 3 (Space group: Cmcm, No. 63). The metal ions have distorted cubical polyhedra with anion similar to the fluorite structure. The excess anions are occupied in the interstitial (empty cubes) of the fluorite unit cell. The crystal structure and chemical analyses suggest approximately equal fractions of U 4+ and U 6+ in this compound. The details of the thermal stability as well as kinetics of formation and oxidation of ZrU 2O 7 are also studied using thermogravimetry.

Aluminum Hydride as a Fuel Supplement to NanoThermites

DTIC Science & Technology

2014-01-01

nanocomposite thermite based on CuO, Bi2O3, and Fe2O3. Pressure cell and burn tube experiments demonstrated enhancements in absolute pressure...pressurization rate, and burning velocity when micron-scale aluminum hydride was used as a minor fuel component in a nanoaluminum–copper-oxide thermite ...alane, AlH3) replaced nanoaluminum incrementally as a fuel in a nanocomposite thermite based on CuO, Bi2O3, and Fe2O3. Pressure cell and burn tube

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

NASA Technical Reports Server (NTRS)

1972-01-01

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

Hydrogen storage systems based on magnesium hydride: from laboratory tests to fuel cell integration

NASA Astrophysics Data System (ADS)

de Rango, P.; Marty, P.; Fruchart, D.

2016-02-01

The paper reviews the state of the art of hydrogen storage systems based on magnesium hydride, emphasizing the role of thermal management, whose effectiveness depends on the effective thermal conductivity of the hydride, but also depends of other limiting factors such as wall contact resistance and convective exchanges with the heat transfer fluid. For daily cycles, the use of phase change material to store the heat of reaction appears to be the most effective solution. The integration with fuel cells (1 kWe proton exchange membrane fuel cell and solid oxide fuel cell) highlights the dynamic behaviour of these systems, which is related to the thermodynamic properties of MgH2. This allows for "self-adaptive" systems that do not require control of the hydrogen flow rate at the inlet of the fuel cell.

Effect of thermo-mechanical processing on microstructure and mechanical properties of U - Nb - Zr alloys: Part 2 - U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr

NASA Astrophysics Data System (ADS)

Morais, Nathanael Wagner Sales; Lopes, Denise Adorno; Schön, Cláudio Geraldo

2018-04-01

The present work is the second and final part of an extended investigation on Usbnd Nb - Zr alloys. It investigates the effect of mechanical processing routes on microstructure of alloys U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr, through X-ray diffraction and scanning electron microscopy, completing the investigation, which started with alloy U - 6 wt% Nb - 6 wt% Zr in part 1. Mechanical properties are determined using microhardness and bending tests and correlated with the developed microstructures. The results show that processing sequence, in particular the inclusion of a 1000 °C heat treatment step, affects significantly the microstructure and mechanical properties of these alloys alloy in different ways. Microstructural characterization shows that both alloys present significant volume fraction of precipitates of a body-centered cubic (BCC) γ-Nb-Zr rich phase in addition the uranium-rich matrix. Bending tests show that sample ductility does not correlate necessarily with hardness and that the key factor appears to be the amount of the γ-Nb-Zr precipitates, which controls the matrix microstructure. Samples with a monoclinic α″ cellular microstructure and/or with the tetragonally-distorted BCC phase (γ0), although not strictly ductile, showed the largest allowed strains-before-break and complete elastic recovery of the broken pieces, pointing out to the macroscopic observation of superelasticity.

Irradiation experiment on ZrC-coated fuel particles for high-temperature gas-cooled reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Minato, Kazuo; Ogawa, Toru; Sawa, Kazuhiro

2000-06-01

The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particle. To compare the irradiation performance of the ZrC Triso-coated fuel particles with that of the normal Triso-coated fuel particles at high temperatures, a capsule irradiation experiment was performed, where both types of the coated fuel particles were irradiated under identical conditions. The burnup was 4.5% FIMA and the irradiation temperature was 1,400 to 1,650 C. The postirradiation measurement of the through-coating failure fractions of both types of coated fuel particles revealed better irradiation performance of the ZrC Triso-coated fuel particles. The opticalmore » microscopy and electron probe microanalysis on the polished cross section of the ZrC Triso-coated fuel particles revealed no interaction of palladium with the ZrC coating layer nor accumulation of palladium at the inner surface of the ZrC coating layer, whereas severe corrosion of the SiC coating layer was observed in the normal Triso-coated fuel particles. Although no corrosion of the ZrC coating layer was observed, additional evaluations need to be made of this layer's ability to satisfactorily retain the fission product palladium.« less

Design and Development of a 30 Watt Solid Polymer Electrolyte Fuel Cell Power Source Fueled with Calcium Hydride.

DTIC Science & Technology

1978-12-12

hydri de and its integration with the fuel cell. I The combination of the SPE cel l with a hydride fuel offers -- comparedto batteries -- increased...demand changes without intermediate storage of hydrogen gas. In order to control the reacti on with water the hydri de is contained in a cartridge. The use

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Capture of Hydrogen Using ZrNi

NASA Technical Reports Server (NTRS)

Patton, Lisa; Wales, Joshua; Lynch, David; Parrish, Clyde

2005-01-01

Water, as ice, is thought to reside in craters at the lunar poles along with CH4 and H2 . A proposed robotic mission for 2012 will utilize metal/metal hydrides for H2 recovery. Specifications are 99% capture of H2 initially at 5 bar and 100C (or greater), and degassing completely at 300C. Of 47-systems examined using the van't Hoff equation, 4 systems, Mg/MgH2, Mg2Ni/Mg2NiH4, ZrNi/ZrNiH2.8, and Pd/PdH0.77, were considered likely candidates for further examination. It is essential, when selecting a system, to also examine questions regarding activation, kinetics, cyclic stability, and gas impurity effects. After considering those issues, ZrN1 was selected as the most promising candidate, as it is easily activated and rapidly forms ZrNiH 2.8 . In addition, it resists oxide poisoning by CO2, and H2O, while some oxidation by O2 is recommended for improved activation . The presence of hydrogen in the as received Zr-Ni alloy from Alfa Aesar posed additional technical problems. X-ray diffraction of the Zr-Ni powder (-325 mesh), with a Zr:Ni wt% ratio of 70:30, was found to consist of ZrH2, ZrNiH2.8, and ZrNi. ZrH2 in the alloy presented the risk that after degassing that both Zr and ZrNi would be present, and thus lead to erroneous results regarding the reactivity of ZrNi with H2 . Fortunately, ZrH2 is a highly stable hydride that does not degas H2 to any significant extent at temperatures below 300C. Based on equilibrium calculations for the decomposition of ZrH2, only 1 millionth of the hydride decomposed at 300C under a N2 atmosphere flowing at 25 ccm for 64 hours, the longest time for pretreatment employed in the investigation. It was possible, from the X-ray results and knowledge of the Zr:Ni ratio, to compute the composition of a pretreated specimen as being 76 wt% ZrNi and the balance ZrH2.

The potential pyrophoricity of BMI-SPEC and aluminum plate spent fuels retrieved from underwater storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ebner, M.A.

1996-08-01

Physical/chemical factors in U metal and hydride combustion, particularly pyrophoricity in ambient environment, were evaluated for BMI-SPEC and UAl{sub x} plate fuels. Some metal fuels may be highly reactive (spontaneously igniting in air) due to high specific surface area, high decay heat, or a high U hydride content from corrosion during underwater storage. However, for the BMI-SPEC and the aluminum plate fuels, this reactivity is too low to present a realistic threat of uncontrolled spontaneous combustion at ambient conditions. While residual U hydride is expected in these corroded fuels, the hydride levels are expected to be too low and themore » configuration too unfavorable to ignite the fuel meat when the fuels are retrieved from the basin and dried. Furthermore the composition and microstructure of the UAl{sub x} fuels further mitigate that risk.« less

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

Dissolution process for ZrO.sub.2 -UO.sub.2 -CaO fuels

DOEpatents

Paige, Bernice E.

1976-06-22

The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE PAGES

Carmack, W. Jon; Chichester, Heather M.; Porter, Douglas L.; ...

2016-02-27

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This then places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. After comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.

2016-05-01

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peakmore » fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.« less

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Microstructural Characterization of Irradiated U0.7ZrH1.6 Using Ultrasonic Techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramuhalli, Pradeep; Jacob, Richard E.; MacFarlan, Paul J.

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such “traditional”more » examination techniques were recently used to characterize a novel new fuel consisting of U0.17ZrH1.6 pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. However, alternative methods such as ultrasonic inspection can provide an opportunity for nondestructively assessing microstructure in both in-pile and post-irradiation examinations. In this paper, we briefly describe initial results of ultrasonic examination of the U0.17ZrH1.6 pellets (unirradiated and irradiated), in a post-irradiation examination study. Data indicate some correlation with microstructural changes due to irradiation; however, it is not clear what the specific microstructural changes are that are influencing the ultrasonic measurements. Interestingly, specimens with nominally identical burnup show differences in ultrasonic signatures, indicating apparent microstructural differences between these specimens. A summary of the experimental study, preliminary data and findings are presented in this short paper. Additional details of the analysis will be included in the presentation.« less

Assessment of solid/liquid equilibria in the (U, Zr)O2+y system

NASA Astrophysics Data System (ADS)

Mastromarino, S.; Seibert, A.; Hashem, E.; Ciccioli, A.; Prieur, D.; Scheinost, A.; Stohr, S.; Lajarge, P.; Boshoven, J.; Robba, D.; Ernstberger, M.; Bottomley, D.; Manara, D.

2017-10-01

Solid/liquid equilibria in the system UO2sbnd ZrO2 are revisited in this work by laser heating coupled with fast optical thermometry. Phase transition points newly measured under inert gas are in fair agreement with the early measurements performed by Wisnyi et al., in 1957, the only study available in the literature on the whole pseudo-binary system. In addition, a minimum melting point is identified here for compositions near (U0.6Zr0.4)O2+y, around 2800 K. The solidus line is rather flat on a broad range of compositions around the minimum. It increases for compositions closer to the pure end members, up to the melting point of pure UO2 (3130 K) on one side and pure ZrO2 (2970 K) on the other. Solid state phase transitions (cubic-tetragonal-monoclinic) have also been observed in the ZrO2-rich compositions X-ray diffraction. Investigations under 0.3 MPa air (0.063 MPa O2) revealed a significant decrease in the melting points down to 2500 K-2600 K for increasing uranium content (x(UO2)> 0.2). This was found to be related to further oxidation of uranium dioxide, confirmed by X-ray absorption spectroscopy. For example, a typical oxidised corium composition U0.6Zr0.4O2.13 was observed to solidify at a temperature as low as 2493 K. The current results are important for assessing the thermal stability of the system fuel - cladding in an oxide based nuclear reactor, and for simulating the system behaviour during a hypothetical severe accident.

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

Tuning the Oxidation State, Nuclearity, and Chemistry of Uranium Hydrides with Phenylsilane and Temperature: The Case of the Classic Uranium(III) Hydride Complex [(C 5 Me 5) 2U(μ-H)] 2

DOE PAGES

Pagano, Justin K.; Dorhout, Jacquelyn M.; Czerwinski, Kenneth R.; ...

2016-03-18

Here, this work demonstrates that the oxidation state and chemistry of uranium hydrides can be tuned with temperature and the stoichiometry of phenylsilane. The trivalent uranium hydride [(C 5Me 5) 2U–H] x (5) was found to be comprised of an equilibrium mixture of U(III) hydrides in solution at ambient temperature. A single U(III) species can be selectively prepared by treating (C 5Me5)2UMe2 (4) with 2 equiv of phenylsilane at 50 °C. The U(III) system is a potent reducing agent and displayed chemistry distinct from the U(IV) system [(C 5Me 5) 2U(H)(μ-H)] 2 (2), which was harnessed to prepare a varietymore » of organometallic complexes, including (C 5Me 5) 2U(dmpe)(H) (6), and the novel uranium(IV) metallacyclopentadiene complex (C 5Me 5) 2U(C 4Me 4) (11).« less

Crystallographic study of Si and ZrN coated U-Mo atomised particles and of their interaction with al under thermal annealing

NASA Astrophysics Data System (ADS)

Zweifel, T.; Palancher, H.; Leenaers, A.; Bonnin, A.; Honkimaki, V.; Tucoulou, R.; Van Den Berghe, S.; Jungwirth, R.; Charollais, F.; Petry, W.

2013-11-01

A new type of high density fuel is needed for the conversion of research and test reactors from high to lower enriched uranium. The most promising one is a dispersion of atomized uranium-molybdenum (U-Mo) particles in an Al matrix. However, during in-pile irradiation the growth of an interaction layer between the U-Mo and the Al matrix strongly limits the fuel's performance. To improve the in-pile behaviour, the U-Mo particles can be coated with protective layers. The SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) fuel development project consists of the production, irradiation and post-irradiation examination of 2 flat, full-size dispersion fuel plates containing respectively Si and ZrN coated U-Mo atomized powder dispersed in a pure Al matrix. In this paper X-ray diffraction analyses of the Si and ZrN layers after deposition, fuel plate manufacturing and thermal annealing are reported. It was found for the U-Mo particles coated with ZrN (thickness 1 μm), that the layer is crystalline, and exhibits lower density than the theoretical one. Fuel plate manufacturing does not strongly influence these crystallographic features. For the U-Mo particles coated with Si (thickness 0.6 μm), the measurements of the as received material suggest an amorphous state of the deposited layer. Fuel plate manufacturing strongly modifies its composition: Si reacts with the U-Mo particles and the Al matrix to grow U(Al, Si)3 and U3Si5 phases. Finally both coatings have shown excellent performances under thermal treatment by limiting drastically the U-Mo/Al interdiffusion. U(Al,Si)3 with two lattice parameters (4.16 Å and 4.21 Å), A distorted U3Si5 phase. Note that these phases were not present in the U-Mo(Si) powders. These phases are usually found in the Silicon rich diffusion layer (SiRDL) obtained in dispersed fuels (as-manufactured U-Mo/Al(Si) fuel plates [12,3] or annealed UMo(Si)/Al fuel rods [40]) as well as in diffusion couples (U-Mo/Al(Si7) [37-39] or U

Demonstration of femtosecond laser ablation inductively coupled plasma mass spectrometry for uranium isotopic measurements in U-10Mo nuclear fuel foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Havrilla, George Joseph; Gonzalez, Jhanis

2015-06-10

The use of femtosecond laser ablation inductively coupled plasma mass spectrometry was used to demonstrate the feasibility of measuring the isotopic ratio of uranium directly in U-10Mo fuel foils. The measurements were done on both the flat surface and cross sections of bare and Zr clad U-10Mo fuel foil samples. The results for the depleted uranium content measurements were less than 10% of the accepted U235/238 ratio of 0.0020. Sampling was demonstrated for line scans and elemental mapping over large areas. In addition to the U isotopic ratio measurement, the Zr thickness could be measured as well as trace elementalmore » composition if required. A number of interesting features were observed during the feasibility measurements which could provide the basis for further investigation using this methodology. The results demonstrate the feasibility of using fs-LA-ICP-MS for measuring the U isotopic ratio in U-10Mo fuel foils.« less

Quantifying the stress fields due to a delta-hydride precipitate in alpha-Zr matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tummala, Hareesh; Capolungo, Laurent; Tome, Carlos N.

This report is a preliminary study on δ-hydride precipitate in zirconium alloy performed using 3D discrete dislocation dynamics simulations. The ability of dislocations in modifying the largely anisotropic stress fields developed by the hydride particle in a matrix phase is addressed for a specific dimension of the hydride. The influential role of probable dislocation nucleation at the hydride-matrix interface is reported. Dislocation nucleation around a hydride was found to decrease the shear stress (S 13) and also increase the normal stresses inside the hydride. We derive conclusions on the formation of stacks of hydrides in zirconium alloys. The contribution ofmore » mechanical fields due to dislocations was found to have a non-negligible effect on such process.« less

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existingmore » facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)« less

SSH2S: Hydrogen storage in complex hydrides for an auxiliary power unit based on high temperature proton exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Baricco, Marcello; Bang, Mads; Fichtner, Maximilian; Hauback, Bjorn; Linder, Marc; Luetto, Carlo; Moretto, Pietro; Sgroi, Mauro

2017-02-01

The main objective of the SSH2S (Fuel Cell Coupled Solid State Hydrogen Storage Tank) project was to develop a solid state hydrogen storage tank based on complex hydrides and to fully integrate it with a High Temperature Proton Exchange Membrane (HT-PEM) fuel cell stack. A mixed lithium amide/magnesium hydride system was used as the main storage material for the tank, due to its high gravimetric storage capacity and relatively low hydrogen desorption temperature. The mixed lithium amide/magnesium hydride system was coupled with a standard intermetallic compound to take advantage of its capability to release hydrogen at ambient temperature and to ensure a fast start-up of the system. The hydrogen storage tank was designed to feed a 1 kW HT-PEM stack for 2 h to be used for an Auxiliary Power Unit (APU). A full thermal integration was possible thanks to the high operation temperature of the fuel cell and to the relative low temperature (170 °C) for hydrogen release from the mixed lithium amide/magnesium hydride system.

Laser diagnostics for NTP fuel corrosion studies

NASA Technical Reports Server (NTRS)

Wantuck, Paul J.; Butt, D. P.; Sappey, A. D.

1993-01-01

Viewgraphs and explanations on laser diagnostics for nuclear thermal propulsion (NTP) fuel corrosion studies are presented. Topics covered include: NTP fuels; U-Zr-C system corrosion products; planar laser-induced fluorescence (PLIF); utilization of PLIF for corrosion product characterization of nuclear thermal rocket fuel elements under test; ZrC emission spectrum; and PLIF imaging of ZrC plume.

U-Zr alloy: XPS and TEM study of surface passivation

NASA Astrophysics Data System (ADS)

Paukov, M.; Tkach, I.; Huber, F.; Gouder, T.; Cieslar, M.; Drozdenko, D.; Minarik, P.; Havela, L.

2018-05-01

Surface reactivity of Uranium metal is an important factor limiting its practical applications. Bcc alloys of U with various transition metals are much less reactive than pure Uranium. So as to specify the mechanism of surface protection, we have been studying the U-20 at.% Zr alloy by photoelectron spectroscopy and transmission electron microscopy. The surface was studied in as-obtained state, in various stages of surface cleaning, and during an isochronal annealing cycle. The analysis based on U-4f, Zr-3p, and O-1 s spectra shows that a Zr-rich phase segregates at the surface at temperatures exceeding 550 K, which provides a self-assembled coating. The comparison of oxygen exposure of the stoichiometric and coated surfaces shows that the coating is efficiently preventing the oxidation of uranium even at elevated temperatures. The coating can be associated with the UZr2+x phase. TEM study indicated that the coating is about 20 nm thick. For the clean state, the U-4f core-level lines of the bcc alloy are practically identical to those of α-U, revealing similar delocalization of the 5f electronic states.

Phase decomposition of γ-U (bcc) in U-10 wt% Mo fuel alloy during hot isostatic pressing of monolithic fuel plate

NASA Astrophysics Data System (ADS)

Park, Y.; Eriksson, N.; Newell, R.; Keiser, D. D.; Sohn, Y. H.

2016-11-01

Eutectoid decomposition of γ-phase (cI2) into α-phase (oC4) and γ‧-phase (tI6) during the hot isostatic pressing (HIP) of the U-10 wt% Mo (U10Mo) alloy was investigated using monolithic fuel plate samples consisting of U10Mo fuel alloy, Zr diffusion barrier and AA6061 cladding. The decomposition of the γ-phase was observed because the HIP process is carried out near the eutectoid temperature, 555 °C. Initially, a cellular structure, consisting of γ‧-phase surrounded by α-phase, developed from the destabilization of the γ-phase. The cellular structure further developed into an alternating lamellar structure of α- and γ‧-phases. Using scanning electron microscopy and transmission electron microscopy, qualitative and quantitative microstructural analyses were carried out to identify the phase constituents, and elucidate the microstructural development based on time-temperature-transformation diagram of the U10Mo alloy. The destabilization of γ -phase into α- and γ‧-phases would be minimized when HIP process was carried out with rapid ramping/cooling rate and dwell temperature higher than 560 °C.

Modeling of hydride precipitation and re-orientation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tikare, Veena; Weck, Philippe F.; Mitchell, John Anthony

In this report, we present a thermodynamic-­based model of hydride precipitation in Zr-based claddings. The model considers the state of the cladding immediately following drying, after removal from cooling-pools, and presents the evolution of precipitate formation upon cooling as follows: The pilgering process used to form Zr-based cladding imparts strong crystallographic and grain shape texture, with the basal plane of the hexagonal α-Zr grains being strongly aligned in the rolling-­direction and the grains are elongated with grain size being approximately twice as long parallel to the rolling direction, which is also the long axis of the tubular cladding, as itmore » is in the orthogonal directions.« less

Integrated Cabin and Fuel Cell System Thermal Management with a Metal Hydride Heat Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hovland, V.

2004-12-01

Integrated approaches for the heating and cooling requirements of both the fuel cell (FC) stack and cabin environment are critical to fuel cell vehicle performance in terms of stack efficiency, fuel economy, and cost. An integrated FC system and cabin thermal management system would address the cabin cooling and heating requirements, control the temperature of the stack by mitigating the waste heat, and ideally capture the waste heat and use it for useful purposes. Current work at the National Renewable Energy Laboratory (NREL) details a conceptual design of a metal hydride heat pump (MHHP) for the fuel cell system andmore » cabin thermal management.« less

Calcium hydride synthesis of Ti-Nb-based alloy powders

NASA Astrophysics Data System (ADS)

Kasimtsev, A. V.; Shuitsev, A. V.; Yudin, S. N.; Levinskii, Yu. V.; Sviridova, T. A.; Alpatov, A. V.; Novosvetlova, E. E.

2017-09-01

The metallothermic (calcium hydride) synthesis of Ti-Nb alloy powders alloyed with tantalum and zirconium is experimentally studied under various conditions. Chemical, X-ray diffraction, and metallographic analyses of the synthesized products show that initial oxides are completely reduced and a homogeneous β-Ti-based alloy powder forms under the optimum synthesis conditions at a temperature of 1200°C. At a lower synthesis temperature, the end products have a high oxygen content. The experimental results are used to plot the thermokinetic dependences o formation of a bcc solid solution at various times of isothermal holding of Ti-22Nb-6Ta and Ti-22Nb-6Zr (at %) alloys. The physicochemical and technological properties of the Ti-22Nb-6Ta and Ti-22Nb-6Zr alloy powders synthesized by calcium hydride reduction under the optimum conditions are determined.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

Separation of actinides from irradiated An-Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl-KCl

NASA Astrophysics Data System (ADS)

Souček, P.; Murakami, T.; Claux, B.; Meier, R.; Malmbeck, R.; Tsukada, T.; Glatz, J.-P.

2015-04-01

An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl-KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An-Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U67-Pu19-Zr10-MA2-RE2 (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide-aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes.

Metastable Metal Hydrides for Hydrogen Storage

DOE PAGES

Graetz, Jason

2012-01-01

The possibility of using hydrogen as a reliable energy carrier for both stationary and mobile applications has gained renewed interest in recent years due to improvements in high temperature fuel cells and a reduction in hydrogen production costs. However, a number of challenges remain and new media are needed that are capable of safely storing hydrogen with high gravimetric and volumetric densities. Metal hydrides and complex metal hydrides offer some hope of overcoming these challenges; however, many of the high capacity “reversible” hydrides exhibit a large endothermic decomposition enthalpy making it difficult to release the hydrogen at low temperatures. Onmore » the other hand, the metastable hydrides are characterized by a low reaction enthalpy and a decomposition reaction that is thermodynamically favorable under ambient conditions. The rapid, low temperature hydrogen evolution rates that can be achieved with these materials offer much promise for mobile PEM fuel cell applications. However, a critical challenge exists to develop new methods to regenerate these hydrides directly from the reactants and hydrogen gas. This spotlight paper presents an overview of some of the metastable metal hydrides for hydrogen storage and a few new approaches being investigated to address the key challenges associated with these materials.« less

The preparation of Zr-deuteride and phase stability studies of the Zr-D system

NASA Astrophysics Data System (ADS)

Maimaitiyili, T.; Steuwer, A.; Bjerkén, C.; Blomqvist, J.; Hoelzel, M.; Ion, J. C.; Zanellato, O.

2017-03-01

Deuteride phases in the zirconium-deuterium system in the temperature range 25-286 °C have been studied in-situ by high resolution neutron diffraction. The study primarily focused on observations of δ→γ transformation at 180 °C, and the peritectoid reaction α + δ ↔ γ at 255 °C in commercial grade Zr powder that was deuterated to a deuterium/Zr ratio of one to one. A detailed description of the zirconium deuteride preparation route by high temperature gas loading is also described. The lattice parameters of α-Zr, δ-ZrDx and ε-ZrDx were determined by whole pattern crystal structure analysis, using Rietveld and Pawley refinements, and are in good agreement with values reported in the literature. The controversial γ-hydride phase was observed both in-situ and ex-situ in deuterated Zr powder after a heat treatment at 286 °C and slow cooling.

Irradiation testing of high density uranium alloy dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

Metal hydride and pyrophoric fuel additives for dicyclopentadiene based hybrid propellants

NASA Astrophysics Data System (ADS)

Shark, Steven C.

The purpose of this study is to investigate the use of reactive energetic fuel additives that have the potential to increase the combustion performance of hybrid rocket propellants in terms of solid fuel regression rate and combustion efficiency. Additives that can augment the combustion flame zone in a hybrid rocket motor by means of increased energy feedback to the fuel grain surface are of great interest. Metal hydrides have large volumetric hydrogen densities, which gives these materials high performance potential as fuel additives in terms of specifc impulse. The excess hydrogen and corresponding base metal may also cause an increase in the hybrid rocket solid fuel regression rate. Pyrophoric additives also have potential to increase the solid fuel regression rate by reacting more readily near the burning fuel surface providing rapid energy feedback. An experimental performance evaluation of metal hydride fuel additives for hybrid rocket motor propulsion systems is examined in this study. Hypergolic ignition droplet tests and an accelerated aging study revealed the protection capabilities of Dicyclopentadiene (DCPD) as a fuel binder, and the ability for unaided ignition. Static hybrid rocket motor experiments were conducted using DCPD as the fuel. Sodium borohydride (NabH4) and aluminum hydride (AlH3) were examined as fuel additives. Ninety percent rocket grade hydrogen peroxide (RGHP) was used as the oxidizer. In this study, the sensitivity of solid fuel regression rate and characteristic velocity (C*) efficiency to total fuel grain port mass flux and particle loading is examined. These results were compared to HTPB combustion performance as a baseline. Chamber pressure histories revealed steady motor operation in most tests, with reduced ignition delays when using NabH4 as a fuel additive. The addition of NabH4 and AlH3 produced up to a 47% and 85% increase in regression rate over neat DCPD, respectively. For all test conditions examined C* efficiency ranges

ORNL Interim Progress Report on Hydride Reorientation CIRFT Tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Yan, Yong; Wang, Hong

A systematic study of H. B. Robinson (HBR) high burnup spent nuclear fuel (SNF) vibration integrity was performed in Phase I project under simulated transportation environments, using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot cell testing technology developed at Oak Ridge National Laboratory in 2013–14. The data analysis on the as-irradiated HBR SNF rods demonstrated that the load amplitude is the dominant factor that controls the fatigue life of bending rods. However, previous studies have shown that the hydrogen content and hydride morphology has an important effect on zirconium alloy mechanical properties. To address the effect of radial hydridesmore » in SNF rods, in Phase II a test procedure was developed to simulate the effects of elevated temperatures, pressures, and stresses during transfer-drying operations. Pressurized and sealed fuel segments were heated to the target temperature for a preset hold time and slow-cooled at a controlled rate. The procedure was applied to both non-irradiated/prehydrided and high-burnup Zircaloy-4 fueled cladding segments using the Nuclear Regulatory Commission-recommended 400°C maximum temperature limit at various cooling rates. Before testing high-burnup cladding, four out-of-cell tests were conducted to optimize the hydride reorientation (R) test condition with pre-hydride Zircaloy-4 cladding, which has the same geometry as the high burnup fuel samples. Test HR-HBR#1 was conducted at the maximum hoop stress of 145 MPa, at a 400°C maximum temperature and a 5°C/h cooling rate. On the other hand, thermal cycling was performed for tests HR-HBR#2, HR-HBR#3, and HR-HBR#4 to generate more radial hydrides. It is clear that thermal cycling increases the ratio of the radial hydride to circumferential hydrides. The internal pressure also has a significant effect on the radial hydride morphology. This report describes a procedure and experimental results of the four out-of-cell hydride reorientation

Study of a ;hot; particle with a matrix of U-bearing metallic Zr: Clue to supercriticality during the Chernobyl nuclear accident

NASA Astrophysics Data System (ADS)

Pöml, P.; Burakov, B.

2017-05-01

This paper is dedicated to the 30th anniversary of the severe nuclear accident that occurred at the Chernobyl NPP on 26 April 1986. A detailed study on a Chernobyl "hot" particle collected from contaminated soil was performed. Optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample. The results show that the particle (≈ 240 x 165 μm) consists of a metallic Zr matrix containing 2-3 wt. % U and bearing veins of an U,Nb admixture. The metallic Zr matrix contains two phases with different amounts of O with the atomic proportions (U,Zr,Nb)0.73O0.27 and (U,Zr,Nb)0.61O0.39. The results confirm the interaction between UO2 fuel and zircaloy cladding in the reactor core. To explain the process of formation of the particle, its properties are compared to laboratory experiments. Because of the metallic nature of the particle it is concluded that it must have formed during a very high temperature (> 2400∘C) process that lasted for only a very short time (few microseconds or less); otherwise the particle should have been oxidised. Such a rapid very high temperature process indicates that at least part of the reactor core could have been supercritical prior to an explosion as it was previously suggested in the literature.

Stability Study of the RERTR Fuel Microstructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jian Gan; Dennis Keiser; Brandon Miller

2014-04-01

The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degreesmore » C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.« less

FCRD Advanced Reactor (Transmutation) Fuels Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, importantmore » to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Microstructure studies of interdiffusion behavior of U 3Si 2/Zircaloy-4 at 800 and 1000 °C

DOE PAGES

He, Lingfeng; Harp, Jason M.; Hoggan, Rita E.; ...

2017-01-22

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U 3Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000°C up to 100 hours. The microstructure of pristine U 3Si 2 and U 3Si 2/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800°C is ZrSi 2,more » with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi 2 interface and Fe-Cr-U-Si phases at ZrSi 2/U-Si interface. As a result, the primary interdiffusion products at 1000°C were Zr 2Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U 6Fe.« less

Hydrides of intermetallic compounds with a H/M ratio greater than unity obtained at high hydrogen pressures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Semenenko, K.N.; Klyamkin, S.N.

1993-11-01

Novel hydride phases with H/M > 1 based on Zr{sub 2}Pd, Hf{sub 2}Pd, and Hf{sub 2}Cu (structures of the MoSi{sub 2} type) have been synthesized at high H{sub 2} pressures. The X-ray diffraction investigations of the resulting hydrides have been carried out. Some factors determining the maximum hydrogen content in the hydrides of intermetallic compounds are discussed. A model structure of the hydrides obtained is proposed, which assumes the possibility of direct H-H interactions when the interatomic distances are less than 1 {angstrom}.

Performance modeling of Deep Burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong

2010-01-01

The effects of design choices for the TRISO particle fuel were explored in order to determine their contribution to attaining high-burnup in Deep Burn modular helium reactor fuels containing transuranics from light water reactor spent fuel. The new design features were: (1) ZrC coating substituted for the SiC, allowing the fuel to survive higher accident temperatures; (2) pyrocarbon/SiC "alloy" substituted for the inner pyrocarbon coating to reduce layer failure and (3) pyrocarbon seal coat and thin ZrC oxygen getter coating on the kernel to eliminate CO. Fuel performance was evaluated using General Atomics Company's PISA code. The only acceptable design has a 200-μm kernel diameter coupled with at least 150-μm thick, 50% porosity buffer, a 15-μm ZrC getter over a 10-μm pyrocarbon seal coat on the kernel, an alloy inner pyrocarbon, and ZrC substituted for SiC. The code predicted that during a 1600 °C postulated accident at 70% FIMA, the ZrC failure probability is <10-4.

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffractionmore » data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results

Modeling and Simulation of Used Nuclear Fuel During Transportation with Consideration of Hydride Effects and Cyclic Fatigue

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chakraborty, Pritam; Sabharwall, Piyush; Spears, Robert Edward

2015-09-30

The objective of this work is to understand the integrity of Used Nuclear Fuel (UNF) during transportation. Previous analysis work has been performed to look at the integrity of UNF during transportation but these analyses have neglected to analyze the effect of hydrides and flaws (fracture mechanics models to capture radial cracking in the cladding). In this study, the clad regions of interest are near the pellet-pellet interfaces. These regions can experience more complex stress-states than the rest of the clad during cooling and have a greater possibility to develop radially reoriented hydrides during vacuum drying.

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

PubMed

Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

2017-07-25

In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Transmission electron microscopy investigation of neutron irradiated Si and ZrN coated UMo particles prepared using FIB

NASA Astrophysics Data System (ADS)

Van Renterghem, W.; Miller, B. D.; Leenaers, A.; Van den Berghe, S.; Gan, J.; Madden, J. W.; Keiser, D. D.

2018-01-01

Two fuel plates, containing Si and ZrN coated U-Mo fuel particles dispersed in an Al matrix, were irradiated in the BR2 reactor of SCK•CEN to a burn-up of ∼70% 235U. Five samples were prepared by INL using focused ion beam milling and transported to SCK•CEN for transmission electron microscopy (TEM) investigation. Two samples were taken from the Si coated U-Mo fuel particles at a burn-up of ∼42% and ∼66% 235U and three samples from the ZrN coated U-Mo at a burn-up of ∼42%, ∼52% and ∼66% 235U. The evolution of the coating, fuel structure, fission products and the formation of interaction layers are discussed. Both coatings appear to be an effective barrier against fuel matrix interaction and only on the samples having received the highest burn-up and power, the formation of an interaction between Al and U(Mo) can be observed on those locations where breaches in the coatings were formed during plate fabrication.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as

Rechargeable metal hydrides for spacecraft application

NASA Technical Reports Server (NTRS)

Perry, J. L.

1988-01-01

Storing hydrogen on board the Space Station presents both safety and logistics problems. Conventional storage using pressurized bottles requires large masses, pressures, and volumes to handle the hydrogen to be used in experiments in the U.S. Laboratory Module and residual hydrogen generated by the ECLSS. Rechargeable metal hydrides may be competitive with conventional storage techniques. The basic theory of hydride behavior is presented and the engineering properties of LaNi5 are discussed to gain a clear understanding of the potential of metal hydrides for handling spacecraft hydrogen resources. Applications to Space Station and the safety of metal hydrides are presented and compared to conventional hydride storage. This comparison indicates that metal hydrides may be safer and require lower pressures, less volume, and less mass to store an equivalent mass of hydrogen.

Hydrogen storage in the form of metal hydrides

NASA Technical Reports Server (NTRS)

Zwanziger, M. G.; Santana, C. C.; Santos, S. C.

1984-01-01

Reversible reactions between hydrogen and such materials as iron/titanium and magnesium/ nickel alloy may provide a means for storing hydrogen fuel. A demonstration model of an iron/titanium hydride storage bed is described. Hydrogen from the hydride storage bed powers a converted gasoline electric generator.

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

NASA Astrophysics Data System (ADS)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

Post Irradiation TEM Investigation of ZrN Coated U(Mo) Particles Prepared with FIB

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Renterghem, W.; Leenaers, A.; Van den Berghe, S.

2015-10-01

In the framework of the Selenium project, two dispersion fuel plates were fabricated with Si and ZrN coated fuel particles and irradiated in the Br2 reactor of SCK•CEN to high burn-up. The first analysis of the irradiated plate proved the reduced swelling of the fuel plate and interaction layer growth up to 70% burn-up. The question was raised how the structure of the interaction layer had been affected by the irradiation and how the structure of the fuel particles had evolved. Hereto, samples from the ZrN coated UMo particles were prepared for transmission electron microscopy (TEM) using focused ion beammore » milling (FIB) at INL. The FIB technique allowed to precisely select the area of the interaction layer and/or fuel to produce a sample that is TEM transparent over an area of 20 by 20 µm. In this contribution, the first TEM results will be presented from the 66% burn-up sample.« less

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

The effect of stress state on zirconium hydride reorientation

NASA Astrophysics Data System (ADS)

Cinbiz, Mahmut Nedim

Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multiaxial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading. Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation. To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to "near-equibiaxial" tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by

Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabin, B. H.; Lloyd, W. R.; Schulthess, J. L.

2015-03-01

This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh)more » fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm 3 to 6.0 x 1021 fissions/cm 3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.« less

Influence of uranium hydride oxidation on uranium metal behaviour

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patel, N.; Hambley, D.; Clarke, S.A.

2013-07-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, ifmore » sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)« less

A DFT+U study of Pu immobilization in Gd2Zr2O7

NASA Astrophysics Data System (ADS)

Zhao, F. A.; Xiao, H. Y.; Jiang, M.; Liu, Z. J.; Zu, X. T.

2015-12-01

The solubility of Pu in Gd2Zr2O7 has been investigated by the density functional theory plus Hubbard U correction. It is found that the formation of PuGdZr2O7, Gd2PuZrO7 and Gd2Pu1.5Zr0.5O7 are exothermic, whereas Pu0.5Gd1.5Zr2O7, Pu1.5Gd0.5Zr2O7 and Gd2Pu0.5Zr1.5O7 are energetically less stable than their respective separated states. The calculations show that both the Gd and Zr lattice sites can be substituted by the Pu, which is consistent with the immobilization behavior of uranium in Gd2Zr2O7 observed experimentally. The site preference of Pu in Gd2Zr2O7 is found to be dependent on the chemical environment, i.e., Pu prefers to substitute for Gd-site under Gd-rich and O2-rich conditions and for Zr-site under Zr-rich and O2-rich conditions.

Selection of Nuclear Fuel for TREAT: UO 2 vs U 3O 8

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glazoff, Michael Vasily; Van Rooyen, Isabella Johanna; Coryell, Benjamin David

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). Themore » TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO 2 vs U 3O 8. In this report, the results are documented pertaining to the choice mentioned above (UO 2 vs U 3O 8). The conclusion in favor of using UO 2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and

Sodium-based hydrides for thermal energy applications

NASA Astrophysics Data System (ADS)

Sheppard, D. A.; Humphries, T. D.; Buckley, C. E.

2016-04-01

Concentrating solar-thermal power (CSP) with thermal energy storage (TES) represents an attractive alternative to conventional fossil fuels for base-load power generation. Sodium alanate (NaAlH4) is a well-known sodium-based complex metal hydride but, more recently, high-temperature sodium-based complex metal hydrides have been considered for TES. This review considers the current state of the art for NaH, NaMgH3- x F x , Na-based transition metal hydrides, NaBH4 and Na3AlH6 for TES and heat pumping applications. These metal hydrides have a number of advantages over other classes of heat storage materials such as high thermal energy storage capacity, low volume, relatively low cost and a wide range of operating temperatures (100 °C to more than 650 °C). Potential safety issues associated with the use of high-temperature sodium-based hydrides are also addressed.

Fundamental experiments on hydride reorientation in zircaloy

NASA Astrophysics Data System (ADS)

Colas, Kimberly B.

reoriented hydride fraction and connectivity increase with number of cycles which could lead to more dangerous microstructure for storage of spent fuel. Pre-existing cracks were also found to affect hydride connectivity and morphology which directly impacts DHC and fuel integrity. (Abstract shortened by UMI.).

Zirconocene-iridium hydrido complexes: arene carbon-hydrogen bond activation and formation of a planar square Zr2Ir2 complex.

PubMed

Oishi, Masataka; Suzuki, Hiroharu

2009-03-16

New early-late heterobimetallic hydrides (L(2)ZrCl)(Cp*Ir)(mu-H)(3) (1; L = Cp derivative, Cp* = eta(5)-C(5)Me(5)) were synthesized from zirconocene derivatives (L(2)ZrCl(2)) and LiCp*IrH(3) via a salt elimination reaction and structurally characterized by NMR and X-ray analyses. Upon treatment of 1 with an alkyllithium reagent, hydride abstraction complex 4 underwent thermolytic ligand elimination at the Zr-Ir system to yield a novel planar square complex (L(2)Zr)(2)(Cp*Ir)(2)(mu(3)-H)(4) (2). When a labeling study of the reaction was conducted, it was found that the conversion of 1 to 2 involves rapid aromatic and benzylic C-H activation by a coordinatively unsaturated dinuclear complex (L(2)Zr)(Cp*Ir)(H)(2) (3).

FY 2016 Status Report: CIRFT Testing on Spent Nuclear Fuels and Hydride Reorientation Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Yan, Yong

This report provides a detailed description of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) testing conducted on spent nuclear fuel (SNF) rods in FY 2016, including hydride reorientation test results. Contact-based measurement, or three-LVDT-based curvature measurement, of SNF rods has proven to be quite reliable in CIRFT testing. However, how the linear variable differential transformer (LVDT) head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. To correct such contact/curvature issues, sensor spacing, defined as the amount of separation between the three LVDT probes, is a criticalmore » measurement that can be used to calculate rod curvature once the deflections are obtained. Recently developed CIRFT data analyses procedures were integrated into FY 2016 CIRFT testing results for the curvature measurements. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNFs into a single zone. A detailed examination revealed that, at same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), Limerick Nuclear Power Station (LMK), mixed uranium-plutonium oxide (MOX). If looking at the strain, then LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The knee point of endurance limit in the curve of moment and curvature or equivalent quantities is more clearly defined for LMK and HBR fuels. The treatment affects the fatigue life of specimens. Both a drop of 12 in. and radial hydride treatment (RHT) have a negative impact on fatigue life. The effect of thermal annealing on MOX fuel rods was relatively small at higher amplitude but became significant at low amplitude of moment. Thermal annealing tended to extend the fatigue

Electron correlation and relativity of the 5f electrons in the U-Zr alloy system

NASA Astrophysics Data System (ADS)

Söderlind, P.; Sadigh, B.; Lordi, V.; Landa, A.; Turchi, P. E. A.

2014-01-01

We address a recently communicated conception that spin-orbit interaction and strong electron correlations are important for the metal fuel U-Zr system. Here, we show that (i) relativistic effects only marginally correct the uranium metal equation-of-state and (ii) addition of onsite Coulomb repulsion leads to an unphysical magnetic ground state of the body-centered cubic (γ) phase and a grossly overestimated equilibrium volume. Consequently, LSDA + U is deemed unsuitable for describing the electronic structure of the U-Zr system. Recently, Xiong et al. [1] reported on thermodynamic modeling of the U-Zr system motivated by its potential as a nuclear fuel for fast breeder reactors. This work [1] came on the heels of another report by Landa et al. [2] on the same system, but with very different results for the formation enthalpies and ultimate conclusion on the U-Zr phase diagram. The authors [1] argue that their calculated energetics are significantly more accurate than that by Landa et al. [2], and they further attribute the difference to strong electron correlations and the relativistic spin-orbit interaction.In the present letter we show that uranium metal, and thus the U-Zr metal nuclear fuel system, possess weakly correlated electrons that are adequately described within density-functional theory in the generalized gradient approximation, and that addition of onsite Coulomb repulsion using the LSDA + U formalism leads to finite magnetization of the γ phase in contradiction to experiments. Furthermore, we show that spin-orbit interaction is quite weak in uranium metal and that its inclusion will not significantly change the chemical bonding and formation enthalpies.In order to illustrate our arguments, we perform comparative electronic-structure calculations using the full-potential linear augmented plane-wave (FPLAPW) method and the projector augmented plane-wave (PAW) method as implemented in the Wien2K [3

Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa

2002-07-01

A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled coremore » has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)« less

FCRD Transmutation Fuels Handbook 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2015-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloysmore » containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about

Effect of hydrogenation conditions on the microstructure and mechanical properties of zirconium hydride

NASA Astrophysics Data System (ADS)

Muta, Hiroaki; Nishikane, Ryoji; Ando, Yusuke; Matsunaga, Junji; Sakamoto, Kan; Harjo, Stefanus; Kawasaki, Takuro; Ohishi, Yuji; Kurosaki, Ken; Yamanaka, Shinsuke

2018-03-01

Precipitation of brittle zirconium hydrides deteriorate the fracture toughness of the fuel cladding tubes of light water reactor. Although the hydride embrittlement has been studied extensively, little is known about physical properties of the hydride due to the experimental difficulties. In the present study, to elucidate relationship between mechanical properties and microstructure, two δ-phase zirconium hydrides and one ε-phase zirconium hydride were carefully fabricated considering volume changes at the metal-to-hydride transformation. The δ-hydride that was fabricated from α-zirconium exhibits numerous inner cracks due to the large volume change. Analyses of the neutron diffraction pattern and electron backscatter diffraction (EBSD) data show that the sample displays significant stacking faults in the {111} plane and in the pseudo-layered microstructure. On the other hand, the δ-hydride sample fabricated from β-zirconium at a higher temperature displays equiaxed grains and no cracks. The strong crystal orientation dependence of mechanical properties were confirmed by indentation test and EBSD observation. The δ-hydride hydrogenated from α-zirconium displays a lower Young's modulus than that prepared from β-zirconium. The difference is attributed to stacking faults within the {111} plane, for which the Young's modulus exhibits the highest value in the perpendicular direction. The strong influence of the crystal orientation and dislocation density on the mechanical properties should be considered when evaluating hydride precipitates in nuclear fuel cladding.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Experimental study of the Er-Zr-O ternary system at 800 °C and 1100 °C

NASA Astrophysics Data System (ADS)

Mascaro, A.; Jourdan, J.; Toffolon-Masclet, C.; Joubert, J.-M.

2012-08-01

The Er-O-Zr ternary system has been investigated experimentally along two isothermal sections at 800 °C and 1100 °C. In order to obtain pure and homogeneous samples, powder metallurgy has been used. The samples have been synthesized using pure Er and Zr powder obtained by the hydride route. The study has been focused on the Zr rich corner and the results allow defining the co-solubility domains at both temperatures and the nature of the phases in equilibrium with αZr and βZr.

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

2014-04-07

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development suchmore » that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.« less

Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motta, Arthur; Ivanov, Kostadin; Arramova, Maria

2015-04-29

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split intomore » two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.« less

The storage of hydrogen in the form of metal hydrides: An application to thermal engines

NASA Technical Reports Server (NTRS)

Gales, C.; Perroud, P.

1981-01-01

The possibility of using LaNi56, FeTiH2, or MgH2 as metal hydride storage sytems for hydrogen fueled automobile engines is discussed. Magnesium copper and magnesium nickel hydrides studies indicate that they provide more stable storage systems than pure magnesium hydrides. Several test engines employing hydrogen fuel have been developed: a single cylinder motor originally designed for use with air gasoline mixture; a four-cylinder engine modified to run on an air hydrogen mixture; and a gas turbine.

Development of a new casting method to fabricate U–Zr alloy containing minor actinides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jong Hwan Kim; Hoon Song; Hyung Tae Kim

2014-01-01

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U–10 wt% Zr, U–10 wt% Zr–5 wt% RE, and U–10 wt% Zr–5 wt% RE–5 wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxyen,more » carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere.« less

Oxidation of U-20 at% Zr alloy in air at 423 1063 K

NASA Astrophysics Data System (ADS)

Matsui, Tsuneo; Yamada, Takanobu; Ikai, Yasushi; Naito, Keiji

1993-01-01

The oxidation behavior of U 0.80Zr 0.20 alloy (two-phase mixture of U and UZr 2 below 878 K and single solid solution above 1008 K) was studied by thermogravimetry in the temperature range from 423 to 1063 K in air. During oxidation in the low temperature region (423-503 K), the sample kept its initial shape (a rectangular rod) and the surface of the sample was covered by a black thin adherent UO2 + x oxide layer. On the other hand, by oxidation in the middle temperature region, the sample broke to several pieces of thin plates and blocks, and fine powder at 643-723 K and entirely to fine powder at 775-878 K, all of which were analyzed to be a mixture of U 3O 8 and ZrO 2. By oxidation in the high temperature region (1008-1063 K) the sample broke to very fine powder, which consisted of U 3O 8 and ZrO 2. Based on the sample shape, the oxide phase identified after oxidation and the slope value of the bilogarithmic plots of the weight gain against time, the oxidation kinetics was analyzed with a paralinear equation in the low temperature region below 503 K and a linear equation in the middle and high temperature regions above 643 K. Oxidation rates of U 0.80Zr 0.20 (two-phase mixture) in the low and middle temperature regions were smaller than those of uranium metal. A discontinuity in the plot of the linear oxidation rate constant versus reciprocal temperature was found to be present between 723 and 838 K, similarly to the case of uranium metal previously reported. The linear rate constants of single-phase solid solution in the high temperature region above 1008 K seemed to be a little smaller than those estimated by the extrapolation of the values in the middle temperature region.

Purification of nuclear grade Zr scrap as the high purity dense Zr deposits from Zirlo scrap by electrorefining in LiF-KF-ZrF4 molten fluorides

NASA Astrophysics Data System (ADS)

Park, Kyoung Tae; Lee, Tae Hyuk; Jo, Nam Chan; Nersisyan, Hayk H.; Chun, Byong Sun; Lee, Hyuk Hee; Lee, Jong Hyeon

2013-05-01

Zirconium (Zr) has commonly been used as a cladding material of nuclear fuel. Moreover, it is regarded as the only material that can be used for nuclear fuel cladding because it has the lowest neutron capture cross section of any metal element and because it has high corrosion resistance and size stability. In this study, Hf-free Zr tubes (Zr-1Nb-1Sn-0.1Fe) were used as anode materials and electrorefining was performed in a LiF-KF eutectic 6 wt.% ZrF4 molten fluoride salt system. As a result of electrolysis, Zr scrap metal was recycled into pure Zr with low levels of impurities, and the size and density of the Zr deposit was controlled using applied current density.

A study of H+ production using metal hydride and other compounds by means of laser ion source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sekine M.; Kondo K.; Okamura, M.

2012-02-22

A laser ion source can provide wide variety of ion beams from solid target materials, however, it has been difficult to create proton beam efficiently. We examined capability of proton production using beeswax, polyethylene, and metal hydrides (MgH2 and ZrH2) as target materials. The results showed that beeswax and polyethylene could not be used to produce protons because these targets are transparent to the laser wavelength of 1064 nm. On the other hand, the metal hydrides could supply protons. Although the obtained particle numbers of protons were less than those of the metal ions, the metal hydrides could be usedmore » as a target for proton laser ion source.« less

Technical and economic aspects of hydrogen storage in metal hydrides

NASA Technical Reports Server (NTRS)

Schmitt, R.

1981-01-01

The recovery of hydrogen from such metal hydrides as LiH, MgH2, TiH2, CaH2 and FeTiH compounds is studied, with the aim of evaluating the viability of the technique for the storage of hydrogen fuel. The pressure-temperature dependence of the reactions, enthalpies of formation, the kinetics of the hydrogen absorption and desorption, and the mechanical and chemical stability of the metal hydrides are taken into account in the evaluation. Economic aspects are considered. Development of portable metal hydride hydrogen storage reservoirs is also mentioned.

Sintering behavior of U 80 at.%Zr powder compacts in a vacuum environment

NASA Astrophysics Data System (ADS)

Kim, Tae-Kyu; Lee, Chong-Tak; Sohn, Dong-Seong

2008-01-01

Sintering behavior of U-80 at.%Zr powder compacts in a temperature range from 1100 to 1500 °C in a vacuum of 1 × 10 -4 Pa was evaluated. The sintered density depended more on the sintering temperature than on the sintering time. The sintered specimens consisted of the δ-UZr 2 matrix with acicular α-Zr precipitates, but it still had un-reacted zirconium when the sintering temperature was 1100 °C. The uranium depletion near the surface of the specimens sintered at temperatures above 1300 °C was detected. Massive Zr(O) grains in the sintered specimen were found, and their formation was restrained when the cooling rate from the sintering temperature was increased.

Ab-initio study of high temperature lattice dynamics of BCC zirconium (β-Zr) and uranium (γ-U)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghosh, Partha S., E-mail: parthasarathi13@gmail.com; Arya, A., E-mail: parthasarathi13@gmail.com; Dey, G. K., E-mail: parthasarathi13@gmail.com

2014-04-24

Using self consistent ab-initio lattice dynamics calculations, we show that bcc structures of Zr and U phases become stable at high temperature by phonon-phonon interactions. The calculated temperature dependent phonon dispersion curve (PDC) of β-Zr match excellently with experimental PDC. But the calculated PDC for γ-U shows negative phonon frequencies even at solid to liquid transition temperature. We show that this discrepancy is due to an overestimation of instability depth of bcc U phase which is removed by incorporation of spin-orbit coupling in the electronic structure calculations.

Process for massively hydriding zirconium--uranium fuel elements

DOEpatents

Katz, N.H.

1973-12-01

A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Matos, J.E.

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersionmore » fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}.« less

Chemical Hydride Slurry for Hydrogen Production and Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClaine, Andrew W

2008-09-30

have demonstrated the technical viability of the process and have provided data for the cost analyses that have been performed. We also concluded that a carbothermic process could also produce magnesium at acceptable costs. The use of slurry as a medium to carry chemical hydrides has been shown during this project to offer significant advantages for storing, delivering, and distributing hydrogen: • Magnesium hydride slurry is stable for months and pumpable. • The oils of the slurry minimize the contact of oxygen and moisture in the air with the metal hydride in the slurry. Thus reactive chemicals, such as lithium hydride, can be handled safely in the air when encased in the oils of the slurry. • Though magnesium hydride offers an additional safety feature of not reacting readily with water at room temperatures, it does react readily with water at temperatures above the boiling point of water. Thus when hydrogen is needed, the slurry and water are heated until the reaction begins, then the reaction energy provides heat for more slurry and water to be heated. • The reaction system can be relatively small and light and the slurry can be stored in conventional liquid fuel tanks. When transported and stored, the conventional liquid fuel infrastructure can be used. • The particular metal hydride of interest in this project, magnesium hydride, forms benign byproducts, magnesium hydroxide (“Milk of Magnesia”) and magnesium oxide. • We have estimated that a magnesium hydride slurry system (including the mixer device and tanks) could meet the DOE 2010 energy density goals. During the investigation of hydriding techniques, we learned that magnesium hydride in a slurry can also be cycled in a rechargeable fashion. Thus, magnesium hydride slurry can act either as a chemical hydride storage medium or as a rechargeable hydride storage system. Hydrogen can be stored and delivered and then stored again thus significantly reducing the cost of storing and delivering

A study of binuclear zirconium hydride catalysts of the hydrogenolysis of alkanes by the density functional theory method

NASA Astrophysics Data System (ADS)

Ustynyuk, L. Yu.; Fast, A. S.; Ustynyuk, Yu. A.; Lunin, V. V.

2012-06-01

Binuclear hydride centers containing two Zr(IV) atoms are suggested as promising catalysts for the hydrogenolysis of alkanes under mild conditions ( T < 450 K, p ˜ 1 atm). Reactions of model compounds L2(H)Zr(X)2Zr(H)L2 (X = H, L = OSi≡ ( 4a), X = L = OMe ( 4d)), L(H)Zr(O)2Zr(H)L (L = OSi≡ ( 4b), Cp( 4c)) and (≡SiO)2(H)Zr-O-Zr(H)(OSi≡)2 ( 4e and 4f) with the propane molecule were studied using the density functional theory method. The results show that centers of the 4a, 4e, and 4f types and especially 4b are promising catalysts of the hydrogenolysis of alkanes due to a high degree of unsaturation of two Zr atoms and their sequential participation in the splitting of the C-C bond and hydrogenation of ethylene formed as a result of splitting.

Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rosenbaum, H.S.

1979-03-01

This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCImore » far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2.« less

Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

NASA Astrophysics Data System (ADS)

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

2016-12-01

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ye, B.; Hofman, G. L.; Leenaers, A.

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate themore » updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.« less

The Planck Sorption Cooler: Using Metal Hydrides to Produce 20 K

NASA Technical Reports Server (NTRS)

Pearson, David P.; Bowman, R.; Prina, M.; Wilson, P.

2006-01-01

The Jet Propulsion Laboratory has built and delivered two continuous closed cycle hydrogen Joule-Thomson (JT) cryocoolers for the ESA Planck mission, which will measure the anisotropy in the cosmic microwave background. The metal hydride compressor consists of six sorbent beds containing LaNi4.78Sn0.22 alloy and a low pressure storage bed of the same material. Each sorbent bed contains a separate gas-gap heat switch that couples or isolates the bed with radiators during the compressor operating cycle. ZrNiHx hydride is used in this heat switch. The Planck compressor produces hydrogen gas at a pressure of 48 Bar by heating the hydride to approx.450 K. This gas passes through a cryogenic cold end consisting of a tube-in-tube heat exchanger, three pre-cooling stages to bring the gas to nominally 52 K, a JT value to expand the gas into the two-phase regime at approx.20 K, and two liquid - vapor heat exchangers that must remove 190 and 646 mW of heat respectively.

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Ti12.5Zr21V10Cr8.5MnxCo1.5Ni46.5-x AB2-type metal hydride alloys for electrochemical storage application: Part 1. Structural characteristics

NASA Astrophysics Data System (ADS)

Bendersky, L. A.; Wang, K.; Levin, I.; Newbury, D.; Young, K.; Chao, B.; Creuziger, A.

2012-11-01

The microstructures of a series of AB2-based metal hydride alloys (Ti12.5Zr21V10Cr8.5MnxCo1.5Ni46.5-x) designed to have different fractions of non-Laves secondary phases were studied by X-ray diffraction, scanning electron microscopy, transmission electron microscopy, energy dispersive X-ray spectrometry, and electron backscatter diffraction. The results indicate that the alloys contain a majority of hydrogen storage Laves phases and a minority of fine-structured non-Laves phases. Formation of the phases is accomplished by dendritic growth of a hexagonal C14 Laves phase. The C14 phase is followed by either a peritectic solidification of a cubic C15 Laves phase (low Mn containing alloys) or a C14 phase of different composition (high Mn containing alloys), and finally a B2 phase formed in the interdendritic regions (IDR). The interdendritic regions may then undergo further solid-state transformation into Zr7Ni10-type, Zr9Ni11-type and TiNi-type phases. As the Mn content in the alloy increases, the fraction of the C14 phase increases, whereas the fraction of C15 decreases. In the IDRs when the alloy's Mn content increases the Zr9Ni11 phases and Zr7Ni10 phase fraction first increases and then decreases, while the TiNi-based phase fraction first increases and then stabilized at 0.02. IDR compositions can be generally expressed as (Ti,Zr,V,Cr,Mn,Co)50Ni50, which accounted for 7-10% of the overall alloy volume fraction.

Heavy-ion irradiation effects on U3O8 incorporated Gd2Zr2O7 waste forms.

PubMed

Lu, Xirui; Shu, Xiaoyan; Chen, Shunzhang; Zhang, Kuibao; Chi, Fangtin; Zhang, Haibin; Shao, Dadong; Mao, Xueli

2018-06-12

In this research, the heavy-ion irradiation effects of U-bearing Gd 2 Zr 2 O 7 ceramics were explored for nuclear waste immobilization. U 3 O 8 was designed to be incorporated into Gd 2 Zr 2 O 7 from two different routes in the form of (Gd 1-4 x U 2 x ) 2 (Zr 1- x U x ) 2 O 7 (x = 0.1, 0.14). The self-irradiation of actinide nuclides was simulated by Xe 20+ heavy-ion radiation under different fluences. Grazing incidence X-ray diffraction (GIXRD) analysis reveals the relationship between radiation dose, damage and depth. The radiation tolerance is promoted with the increment of U 3 O 8 content in the discussed range. Raman spectroscopy testifies the enhancement of radiation tolerance and microscopically existed phase evolution from the chemical bond vibrations. In addition, the microstructure and elemental distribution of the irradiated samples were analyzed as well. The amorphization degree of the sample surface declines as the U content was elevated from x = 0.1 to x = 0.14. Copyright © 2018 Elsevier B.V. All rights reserved.

Development of Hydrogen Storage Tank Systems Based on Complex Metal Hydrides

PubMed Central

Ley, Morten B.; Meggouh, Mariem; Moury, Romain; Peinecke, Kateryna; Felderhoff, Michael

2015-01-01

This review describes recent research in the development of tank systems based on complex metal hydrides for thermolysis and hydrolysis. Commercial applications using complex metal hydrides are limited, especially for thermolysis-based systems where so far only demonstration projects have been performed. Hydrolysis-based systems find their way in space, naval, military and defense applications due to their compatibility with proton exchange membrane (PEM) fuel cells. Tank design, modeling, and development for thermolysis and hydrolysis systems as well as commercial applications of hydrolysis systems are described in more detail in this review. For thermolysis, mostly sodium aluminum hydride containing tanks were developed, and only a few examples with nitrides, ammonia borane and alane. For hydrolysis, sodium borohydride was the preferred material whereas ammonia borane found less popularity. Recycling of the sodium borohydride spent fuel remains an important part for their commercial viability. PMID:28793541

Laser-induced breakdown spectroscopy of light water reactor simulated used nuclear fuel: Main oxide phase

DOE PAGES

Campbell, Keri R.; Judge, Elizabeth J.; Barefield, James E.; ...

2017-04-22

We show the analysis of light water reactor simulated used nuclear fuel using laser-induced breakdown spectroscopy (LIBS) is explored using a simplified version of the main oxide phase. The main oxide phase consists of the actinides, lanthanides, and zirconium. The purpose of this study is to develop a rapid, quantitative technique for measuring zirconium in a uranium dioxide matrix without the need to dissolve the material. A second set of materials including cerium oxide is also analyzed to determine precision and limit of detection (LOD) using LIBS in a complex matrix. Two types of samples are used in this study:more » binary and ternary oxide pellets. The ternary oxide, (U,Zr,Ce)O 2 pellets used in this study are a simplified version the main oxide phase of used nuclear fuel. The binary oxides, (U,Ce)O 2 and (U,Zr)O 2 are also examined to determine spectral emission lines for Ce and Zr, potential spectral interferences with uranium and baseline LOD values for Ce and Zr in a UO 2 matrix. In the spectral range of 200 to 800 nm, 33 cerium lines and 25 zirconium lines were identified and shown to have linear correlation values (R 2) > 0.97 for both the binary and ternary oxides. The cerium LOD in the (U,Ce)O 2 matrix ranged from 0.34 to 1.08 wt% and 0.94 to 1.22 wt% in (U,Ce,Zr)O 2 for 33 of Ce emission lines. The zirconium limit of detection in the (U,Zr)O 2 matrix ranged from 0.84 to 1.15 wt% and 0.99 to 1.10 wt% in (U,Ce,Zr)O 2 for 25 Zr lines. Finally, the effect of multiple elements in the plasma and the impact on the LOD is discussed.« less

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE PAGES

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

2016-12-01

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Results of NDE Technique Evaluation of Clad Hydrides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kunerth, Dennis C.

2014-09-01

This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used tomore » detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of

The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters

DTIC Science & Technology

2016-01-04

AFRL-AFOSR-VA-TR-2016-0075 The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters KIT BOWEN JOHNS HOPKINS UNIV BALTIMORE MD...Hydride and Boron Aluminum Hydride Clusters 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA9550-14-1-0324 5c.  PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR(S) KIT...of both Aluminum Hydride Cluster Anions and Boron Aluminum Hydride Cluster Anions with Oxygen: Anionic Products The anionic products of reactions

Thermodynamic modelling of the C-U and B-U binary systems

NASA Astrophysics Data System (ADS)

Chevalier, P. Y.; Fischer, E.

2001-02-01

The thermodynamic modelling of the carbon-uranium (C-U) and boron-uranium (B-U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium-concrete interaction (MCCI). The key O-U-Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe-U, Cr-U and Ni-U were modelled as a preliminary work for modelling the O-U-Zr-Fe-Cr-Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver-indium-cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B 4C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B 2O 3 and C with the major components of TDBCR, O-U-Zr-Fe-Cr-Ni-Ag-In-Ba-La-Ru-Sr-Al-Ca-Mg-Si + Ar-H. The critical assessment of the very numerous experimental information available for the C-U and B-U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.

Hydriding process

DOEpatents

Raymond, J.W.; Taketani, H.

1973-12-01

BS>A method is described for hydriding a body of a Group IV-B metal, preferably zirconium, to produce a crack-free metal-hydride bedy of high hydrogen content by cooling the body at the beta to beta + delta boundary, without further addition of hydrogen, to precipitate a fine-grained delta-phase metal hydride in the beta + delta phase region and then resuming the hydriding, preferably preceded by a reheating step. (Official Gazette)

Metal Hydride Compression

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Terry A.; Bowman, Robert; Smith, Barton

Conventional hydrogen compressors often contribute over half of the cost of hydrogen stations, have poor reliability, and have insufficient flow rates for a mature FCEV market. Fatigue associated with their moving parts including cracking of diaphragms and failure of seal leads to failure in conventional compressors, which is exacerbated by the repeated starts and stops expected at fueling stations. Furthermore, the conventional lubrication of these compressors with oil is generally unacceptable at fueling stations due to potential fuel contamination. Metal hydride (MH) technology offers a very good alternative to both conventional (mechanical) and newly developed (electrochemical, ionic liquid pistons) methodsmore » of hydrogen compression. Advantages of MH compression include simplicity in design and operation, absence of moving parts, compactness, safety and reliability, and the possibility to utilize waste industrial heat to power the compressor. Beyond conventional H2 supplies of pipelines or tanker trucks, another attractive scenario is the on-site generating, pressuring and delivering pure H 2 at pressure (≥ 875 bar) for refueling vehicles at electrolysis, wind, or solar generating production facilities in distributed locations that are too remote or widely distributed for cost effective bulk transport. MH hydrogen compression utilizes a reversible heat-driven interaction of a hydride-forming metal alloy with hydrogen gas to form the MH phase and is a promising process for hydrogen energy applications [1,2]. To deliver hydrogen continuously, each stage of the compressor must consist of multiple MH beds with synchronized hydrogenation & dehydrogenation cycles. Multistage pressurization allows achievement of greater compression ratios using reduced temperature swings compared to single stage compressors. The objectives of this project are to investigate and demonstrate on a laboratory scale a two-stage MH hydrogen (H 2) gas compressor with a feed pressure

A study of hydriding kinetics of metal hydrides using a physically based model

NASA Astrophysics Data System (ADS)

Voskuilen, Tyler G.

The reaction of hydrogen with metals to form metal hydrides has numerous potential energy storage and management applications. The metal hydrogen system has a high volumetric energy density and is often reversible with a high cycle life. The stored hydrogen can be used to produce energy through combustion, reaction in a fuel cell, or electrochemically in metal hydride batteries. The high enthalpy of the metal-hydrogen reaction can also be used for rapid heat removal or delivery. However, improving the often poor gravimetric performance of such systems through the use of lightweight metals usually comes at the cost of reduced reaction rates or the requirement of pressure and temperature conditions far from the desired operating conditions. In this work, a 700 bar Sievert system was developed at the Purdue Hydrogen Systems Laboratory to study the kinetic and thermodynamic behavior of high pressure hydrogen absorption under near-ambient temperatures. This system was used to determine the kinetic and thermodynamic properties of TiCrMn, an intermetallic metal hydride of interest due to its ambient temperature performance for vehicular applications. A commonly studied intermetallic hydride, LaNi5, was also characterized as a base case for the phase field model. The analysis of the data obtained from such a system necessitate the use of specialized techniques to decouple the measured reaction rates from experimental conditions. These techniques were also developed as a part of this work. Finally, a phase field model of metal hydride formation in mass-transport limited interstitial solute reactions based on the regular solution model was developed and compared with measured kinetics of LaNi5 and TiCrMn. This model aided in the identification of key reaction features and was used to verify the proposed technique for the analysis of gas-solid reaction rates determined volumetrically. Additionally, the phase field model provided detailed quantitative predictions of the

Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

NASA Astrophysics Data System (ADS)

Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

2018-02-01

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

NASA Astrophysics Data System (ADS)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

Super Hydrides.

DTIC Science & Technology

1988-03-01

enantioselective synthesis Of the clinically important anti-depressants, (-)Tomoxetine, Fluoxetine (Prozac, Eli Lilly), and Nisoxetine (Scheme 1 ). Schem I a I...Scheme 1 . Another salient feature of this synthesis is that it correlated for the first time the absolute configuration of the enantiomers of...RD-RI93 710 SUPER HYDRIDES(U) PURDUE UNIV LRFRYETTE IN H C BROWN 1 / 1 NAR 88 RRO-22302.2-CN DAR29-05-K-1662 UNCLSSIFIED F/G 7/3 NI. t2S 16, L,. 10 3

Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

DOEpatents

Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

2014-11-18

An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

Microstructure and hydrogenation properties of a melt-spun non-stoichiometric Zr-based Laves phase alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Tiebang, E-mail: tiebangzhang@nwpu.edu.cn; Zhang, Yunlong; Li, Jinshan

2016-01-15

Alloy with composition of Zr{sub 0.9}Ti{sub 0.1}V{sub 1.7} off normal stoichiometric proportion is selected to investigate the effect of defects introduced by non-stoichiometry on hydrogenation kinetics of Zr–Ti–V Laves phase alloys. Microstructure and phase constituent of melt-spun ribbons have been investigated in this work. The activation process, hydrogenation kinetics, thermodynamics characteristics and hydride phase constituent of as-cast alloy and melt-spun ribbons are also compared. Comparing with the as-cast alloy, the dominant Laves phase ZrV{sub 2} is preserved, V-BCC phase is reduced and α-Zr phase is replaced by a small amount of Zr{sub 3}V{sub 3}O phase in melt-spun ribbons. Melt-spun ribbonsmore » exhibit easy activation and fast initial hydrogen absorption on account of the increased specific surface area. However, the decrease in unit cell volume of the dominant phase leads to the decrease in hydrogen absorption capacity. Melt-spinning technique raises the equilibrium pressure and decreases the stability of hydride due to the decrease of unit cell volume and the elimination of α-Zr phase, respectively. Melt-spun ribbons with fine grains show improved hydrogen absorption kinetics comparing with that of the as-cast alloy. Meanwhile, the prevalent micro twins observed within melt-spun ribbons are believed to account for the improved hydrogen absorption kinetics. - Highlights: • Role of defects on hydrogenation kinetics of Zr-based alloys is proposed. • Microstructure and hydrogenation properties of as-cast/melt-spun alloy are compared. • Melt-spinning technique improves the hydrogenation kinetics of Zr{sub 0.9}Ti{sub 0.1}V{sub 1.7} alloy. • Refined grains and twin defects account for improved hydrogen absorption kinetics.« less

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Structural Characterization of Metal Hydrides for Energy Applications

NASA Astrophysics Data System (ADS)

George, Lyci

Hydrogen can be an unlimited source of clean energy for future because of its very high energy density compared to the conventional fuels like gasoline. An efficient and safer way of storing hydrogen is in metals and alloys as hydrides. Light metal hydrides, alanates and borohydrides have very good hydrogen storage capacity, but high operation temperatures hinder their application. Improvement of thermodynamic properties of these hydrides is important for their commercial use as a source of energy. Application of pressure on materials can have influence on their properties favoring hydrogen storage. Hydrogen desorption in many complex hydrides occurs above the transition temperature. Therefore, it is important to study the physical properties of the hydride compounds at ambient and high pressure and/or high temperature conditions, which can assist in the design of suitable storage materials with desired thermodynamic properties. The high pressure-temperature phase diagram, thermal expansion and compressibility have only been evaluated for a limited number of hydrides so far. This situation serves as a main motivation for studying such properties of a number of technologically important hydrides. Focus of this dissertation was on X-ray diffraction and Raman spectroscopy studies of Mg2FeH6, Ca(BH4) 2, Mg(BH4)2, NaBH4, NaAlH4, LiAlH4, LiNH2BH3 and mixture of MgH 2 with AlH3 or Si, at different conditions of pressure and temperature, to obtain their bulk modulus and thermal expansion coefficient. These data are potential source of information regarding inter-atomic forces and also serve as a basis for developing theoretical models. Some high pressure phases were identified for the complex hydrides in this study which may have better hydrogen storage properties than the ambient phase. The results showed that the highly compressible B-H or Al-H bonds and the associated bond disordering under pressure is responsible for phase transitions observed in brorohydrides or

Hydrogen Storage Properties of New Hydrogen-Rich BH3NH3-Metal Hydride (TiH2, ZrH2, MgH2, and/or CaH2) Composite Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Young Joon; Xu, Yimin; Shaw, Wendy J.

2012-04-19

Ammonia borane (AB = NH3BH3) is one of the most attractive materials for chemical hydrogen storage due to its high hydrogen contents of 19.6 wt.%, however, impurity levels of borazine, ammonia and diborane in conjunction with foaming and exothermic hydrogen release calls for finding ways to mitigate the decomposition reactions. In this paper we present a solution by mixing AB with metal hydrides (TiH2, ZrH2, MgH2 and CaH2) which have endothermic hydrogen release in order to control the heat release and impurity levels from AB upon decomposition. The composite materials were prepared by mechanical ball milling, and their H2 releasemore » properties were characterized by thermogravimetric analysis (TGA) and differential scanning calorimetry (DSC). The formation of volatile products from decomposition side reactions, such as borazine (N3B3H6) was determined by mass spectrometry (MS). Sieverts type pressure-composition-temperature (PCT) gas-solid reaction instrument was adopted to observe the kinetics of the H2 release reactions of the combined systems and neat AB. In situ 11B MAS-NMR revealed a destabilized decomposition pathway. We found that by adding specific metal hydrides to AB we can eliminate the impurities and mitigate the heat release.« less

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

2014-11-01

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

NASA Astrophysics Data System (ADS)

Serrano-Purroy, D.; Casas, I.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; Clarens, F.; Giménez, J.; de Pablo, J.; Martínez-Esparza, A.

2013-03-01

The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used.For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample.Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample.Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

U.S./CIS eye joint nuclear rocket venture

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

1993-01-01

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

Hydride compositions

DOEpatents

Lee, Myung, W.

1994-01-01

Disclosed are a composition for use in storing hydrogen and a method for making the composition. The composition comprises a mixture of two or more hydrides, each hydride having a different series of hydrogen sorption isotherms that contribute to the overall isotherms of the mixture. The hydrides are chosen so that the isotherms of the mixture have regions wherein the H equilibrium pressure increases with increasing hydrogen, preferably linearly. The isotherms of the mixture can be adjusted by selecting hydrides with different isotherms and by varying the amounts of the individual hydrides, or both. Preferably, the mixture is made up of hydrides that have isotherms with substantially flat plateaus and in nearly equimolar amounts. The composition is activated by degassing, exposing to H, and then heating below the softening temperature of any of the constituents. When the composition is used to store hydrogen, its hydrogen content can be found simply by measuring P{sub H}{sub 2} and determining H/M from the isothermic function of the composition.

Hydride compositions

DOEpatents

Lee, Myung W.

1995-01-01

A composition for use in storing hydrogen, and a method for making the composition. The composition comprises a mixture of two or more hydrides, each hydride having a different series of hydrogen sorption isotherms that contribute to the overall isotherms of the mixture. The hydrides are chosen so that the isotherms of the mixture have regions wherein the hydrogen equilibrium pressure increases with increasing hydrogen, preferably linearly. The isotherms of the mixture can be adjusted by selecting hydrides with different isotherms and by varying the amounts of the individual hydrides, or both. Preferably, the mixture is made up of hydrides that have isotherms with substantially flat plateaus and in nearly equimolar amounts. The composition is activated by degassing, exposing to hydrogen and then heating at a temperature below the softening temperature of any of the. constituents so that their chemical and structural integrity is preserved. When the composition is used to store hydrogen, its hydrogen content can be found simply by measuring P.sub.H.sbsb.2 and determining H/M from the isothermic function of the composition.

Pyroprocess for processing spent nuclear fuel

DOEpatents

Miller, William E.; Tomczuk, Zygmunt

2002-01-01

This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: Zr diffusion barrier with a thickness of 25 μm. A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7-13 wt.%. Decomposed areas containing plate-shaped low-Mo phase. A typical Zr/cladding interaction layer with a thickness of 1-2 μm. A visible UZr2 bearing layer with a thickness of 1-2 μm. Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U-Mo matrix. No excessive interaction between cladding and the uncoated fuel edge. Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. Some of these attributes might be

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

NASA Astrophysics Data System (ADS)

Van Kleeck, Melissa A.

The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

The Current Status of Hydrogen Storage Alloy Development for Electrochemical Applications.

PubMed

Young, Kwo-Hsiung; Nei, Jean

2013-10-17

In this review article, the fundamentals of electrochemical reactions involving metal hydrides are explained, followed by a report of recent progress in hydrogen storage alloys for electrochemical applications. The status of various alloy systems, including AB₅, AB₂, A₂B₇-type, Ti-Ni-based, Mg-Ni-based, BCC, and Zr-Ni-based metal hydride alloys, for their most important electrochemical application, the nickel metal hydride battery, is summarized. Other electrochemical applications, such as Ni-hydrogen, fuel cell, Li-ion battery, air-metal hydride, and hybrid battery systems, also have been mentioned.

Predicting Hydride Donor Strength via Quantum Chemical Calculations of Hydride Transfer Activation Free Energy.

PubMed

Alherz, Abdulaziz; Lim, Chern-Hooi; Hynes, James T; Musgrave, Charles B

2018-01-25

We propose a method to approximate the kinetic properties of hydride donor species by relating the nucleophilicity (N) of a hydride to the activation free energy ΔG ⧧ of its corresponding hydride transfer reaction. N is a kinetic parameter related to the hydride transfer rate constant that quantifies a nucleophilic hydridic species' tendency to donate. Our method estimates N using quantum chemical calculations to compute ΔG ⧧ for hydride transfers from hydride donors to CO 2 in solution. A linear correlation for each class of hydrides is then established between experimentally determined N values and the computationally predicted ΔG ⧧ ; this relationship can then be used to predict nucleophilicity for different hydride donors within each class. This approach is employed to determine N for four different classes of hydride donors: two organic (carbon-based and benzimidazole-based) and two inorganic (boron and silicon) hydride classes. We argue that silicon and boron hydrides are driven by the formation of the more stable Si-O or B-O bond. In contrast, the carbon-based hydrides considered herein are driven by the stability acquired upon rearomatization, a feature making these species of particular interest, because they both exhibit catalytic behavior and can be recycled.

A U-bearing composite waste form for electrochemical processing wastes

NASA Astrophysics Data System (ADS)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

2018-04-01

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases.

The Development of a Compact Refrigeration System using Metal Hydrides

NASA Astrophysics Data System (ADS)

Bae, Sang-Chul; Ogawa, Masahito; Katsuta, Masafumi

The MH refrigeration systems are regarded as important and compact ones for solving energy and environmental issues. Our purposes are to develop the compact refrigeration system for the vending machine and the show case using MH, and to attain a refrigeration temperature of 243K by using a heat source of 403∼423K. The kinetics of MH hydriding and dehydriding reactions is of importance relative to their practical use as a refrigerator system. The kinetics of the reaction between hydrogen and MHHigh (Ti0.18Zr0.84Cr1.0FeO.7Mn0.3CuO.057)has been followed in this paper. A relatively rapid absorption of hydrogen takes place for values of relative composition to about 0.3∼0.4. It is evident that a hydrogen diffusion plays a minor role during this stage, as that part of the metal not covered by hydride is always in contact with hydrogen. The direct chemical reaction between the hydrogen and the exposed metal surface is therefore postulated as the rate-controlling process. The rate of the reaction then decreases, and for values of relative composition above about 0.8, the reaction becomes slow. After the metal particles have been completely covered by a hydride layer, the transport of materials through the layer by diffusion becomes rate controlling process

Assessment of Zr-Fe-V getter alloy for gas-gap heat switches

NASA Technical Reports Server (NTRS)

Prina, M.; Kulleck, J. G.; Bowman, R. C., Jr.

2000-01-01

A commercial Zr-V-Fe alloy (i.e., SAES Getters trade name alloy St-172) has been assessed as reversible hydrogen storage material for use in actuators of gas gap heat switches. Two prototype actuators containing the SAES St-172 material were built and operated for several thousand cycles to evaluate performance of the metal hydride system under conditions simulating heat switch operation.

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

Development of Ni-Ba(Zr,Y)O3 cermet anodes for direct ammonia-fueled solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Miyazaki, Kazunari; Okanishi, Takeou; Muroyama, Hiroki; Matsui, Toshiaki; Eguchi, Koichi

2017-10-01

In this study, the availability of Ni-Ba(Zr,Y)O3-δ (BZY) cermet for the anode of direct ammonia-fueled solid oxide fuel cells (SOFCs) is evaluated. In this device, the anodes need to be active for the catalytic ammonia decomposition as well as the electrochemical hydrogen oxidation. In the catalytic activity test, ammonia decomposes completely over Ni-BZY at ca. 600 °C, while higher temperature is required to accomplish the complete decomposition over the conventional SOFC anode of Ni-yttria-stabilized zirconia cermet. The high activity of Ni-BZY is attributed to the high basicity of BZY and the high resistance to hydrogen poisoning effect. The electrochemical property of Ni-BZY anode is also evaluated with the anode-supported cell of Ni-BZY|BZY|Pt at 600-700 °C with feeding ammonia or hydrogen as a fuel. Since the residence time of ammonia fuel in the thick Ni-BZY anode is long, the difference in the cell performance between two fuels is relatively small. Furthermore, it is proved that the steam concentration in the fuel strongly affects the cell performance. We find that this factor is important to satisfy the above mentioned requirements for the anode of direct ammonia-fueled SOFCs. Throughout this study, it is concluded that Ni-BZY cermet will be a promising anode.

Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

2010-02-23

Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Development of a component design tool for metal hydride heat pumps

NASA Astrophysics Data System (ADS)

Waters, Essene L.

Given current demands for more efficient and environmentally friendly energy sources, hydrogen based energy systems are an increasingly popular field of interest. Within the field, metal hydrides have become a prominent focus of research due to their large hydrogen storage capacity and relative system simplicity and safety. Metal hydride heat pumps constitute one such application, in which heat and hydrogen are transferred to and from metal hydrides. While a significant amount of work has been done to study such systems, the scope of materials selection has been quite limited. Typical studies compare only a few metal hydride materials and provide limited justification for the choice of those few. In this work, a metal hydride component design tool has been developed to enable the targeted down-selection of an extensive database of metal hydrides to identify the most promising materials for use in metal hydride thermal systems. The material database contains over 300 metal hydrides with various physical and thermodynamic properties included for each material. Sub-models for equilibrium pressure, thermophysical data, and default properties are used to predict the behavior of each material within the given system. For a given thermal system, this tool can be used to identify optimal materials out of over 100,000 possible hydride combinations. The selection tool described herein has been applied to a stationary combined heat and power system containing a high-temperature proton exchange membrane (PEM) fuel cell, a hot water tank, and two metal hydride beds used as a heat pump. A variety of factors can be used to select materials including efficiency, maximum and minimum system pressures, pressure difference, coefficient of performance (COP), and COP sensitivity. The targeted down-selection of metal hydrides for this system focuses on the system's COP for each potential pair. The values of COP and COP sensitivity have been used to identify pairs of highest interest for

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Hydride reorientation and its impact on ambient temperature mechanical properties of high burn-up irradiated and unirradiated recrystallized Zircaloy-2 nuclear fuel cladding with an inner liner

NASA Astrophysics Data System (ADS)

Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

2017-10-01

Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually

A U-bearing composite waste form for electrochemical processing wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phasesmore » that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.« less

Simulation of Zr content in TiZrCuNi brazing filler metal for Ti6Al4V alloy

NASA Astrophysics Data System (ADS)

Yue, Xishan; Xie, Zonghong; Jing, Yongjuan

2017-07-01

To optimize the Zr content in Ti-based filler metal, the covalent electron on the nearest atoms bond in unit cell ( n A u-v ) with Ti-based BCC structure was calculated, in which the brazing temperature was considered due to its influence on the lattice parameter. Based on EET theory (The Empirical Electron Theory for solid and molecules), n_{{A}}^{{u - v}} represents the strength of the unit cell with defined element composition and structure, which reflects the effect from solid solution strengthening on the strength of the unit cell. For Ti-Zr-15Cu-10Ni wt% filler metal, it kept constant as 0.3476 with Zr as 37.5˜45 wt% and decreased to 0.333 with Zr decreasing from 37.5 to 25 wt%. Finally, it increased up to 0.3406 with Zr as 2˜10 wt%. Thus, Ti-based filler metal with Zr content being 2˜10 wt% is suggested based on the simulation results. Moreover, the calculated covalent electron of n A u-v showed good agreement with the hardness of the joint by filler 37.5Zr and 10Zr. The composition of Ti-10Zr-15Cu-10Ni wt% was verified in this study with higher tensile strength of the brazing joint and uniform microstructure of the interface.

Thermochemical Assessment of Oxygen Gettering by SiC or ZrC in PuO2-x TRISO Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Besmann, Theodore M

2010-01-01

Particulate nuclear fuel in a modular helium reactor is being considered for the consumption of excess plutonium and related transuranics. In particular, efforts to largely consume transuranics in a single-pass will require the fuel to undergo very high burnup. This deep burn concept will thus make the proposed plutonia TRISO fuel particularly likely to suffer kernel migration where carbon in the buffer layer and inner pyrolytic carbon layer is transported from the high temperature side of the particle to the low temperature side. This phenomenon is oberved to cause particle failure and therefore must be mitigated. The addition of SiCmore » or ZrC in the oxide kernel or in a layer in communication with the kernel will lower the oxygen potential and therefore prevent kernel migration, and this has been demonstrated with SiC. In this work a thermochemical analysis was performed to predict oxygen potential behavior in the plutonia TRISO fuel to burnups of 50% FIMA with and without the presence of oxygen gettering SiC and ZrC. Kernel migration is believed to be controlled by CO gas transporting carbon from the hot side to the cool side, and CO pressure is governed by the oxygen potential in the presence of carbon. The gettering phases significantly reduce the oxygen potential and thus CO pressure in an otherwise PuO2-x kernel, and prevent kernel migration by limiting CO gas diffusion through the buffer layer. The reduction in CO pressure can also reduce the peak pressure within the particles by ~50%, thus reducing the likelihood of pressure-induced particle failure. A model for kernel migration was used to semi-quantitatively assess the effect of controlling oxygen potential with SiC or ZrC and did demonstrated the dramatic effect of the addition of these phases on carbon transport.« less

High temperature reaction between sea salt deposit and (U,Zr)O2 simulated corium debris

NASA Astrophysics Data System (ADS)

Takano, Masahide; Nishi, Tsuyoshi

2013-11-01

In order to clarify the possible impacts of seawater injection on the chemical and physical state of the corium debris formed in the severe accident at Fukushima Daiichi Nuclear Power Plants, the high temperature reaction between sea salt deposit and (U,Zr)O2 simulated corium debris (sim-debris) was examined in the temperature range from 1088 to 1668 K. A dense layer of calcium and sodium uranate formed on the surface of a sim-debris pellet at 1275 K under airflow, with the thickness of over 50 μm. When the oxygen partial pressure is low, calcium is likely to dissolve into the cubic sim-debris phase to form solid solution (Ca,U,Zr)O2+x. The diffusion depth was 5-6 μm from the surface, subjected to 1275 K for 12 h. The crystalline MgO remains affixed on the surface as the main residue of salt components. A part of it can also dissolve into the sim-debris.

The Current Status of Hydrogen Storage Alloy Development for Electrochemical Applications

PubMed Central

Young, Kwo-hsiung; Nei, Jean

2013-01-01

In this review article, the fundamentals of electrochemical reactions involving metal hydrides are explained, followed by a report of recent progress in hydrogen storage alloys for electrochemical applications. The status of various alloy systems, including AB5, AB2, A2B7-type, Ti-Ni-based, Mg-Ni-based, BCC, and Zr-Ni-based metal hydride alloys, for their most important electrochemical application, the nickel metal hydride battery, is summarized. Other electrochemical applications, such as Ni-hydrogen, fuel cell, Li-ion battery, air-metal hydride, and hybrid battery systems, also have been mentioned. PMID:28788349

Hydrogen generation using silicon nanoparticles and their mixtures with alkali metal hydrides

NASA Astrophysics Data System (ADS)

Patki, Gauri Dilip

Hydrogen is a promising energy carrier, for use in fuel cells, engines, and turbines for transportation or mobile applications. Hydrogen is desirable as an energy carrier, because its oxidation by air releases substantial energy (thermally or electrochemically) and produces only water as a product. In contrast, hydrocarbon energy carriers inevitably produce CO2, contributing to global warming. While CO2 capture may prove feasible in large stationary applications, implementing it in transportation and mobile applications is a daunting challenge. Thus a zero-emission energy carrier like hydrogen is especially needed in these cases. Use of H2 as an energy carrier also brings new challenges such as safe handling of compressed hydrogen and implementation of new transport, storage, and delivery processes and infrastructure. With current storage technologies, hydrogen's energy per volume is very low compared to other automobile fuels. High density storage of compressed hydrogen requires combinations of high pressure and/or low temperature that are not very practical. An alternative for storage is use of solid light weight hydrogenous material systems which have long durability, good adsorption properties and high activity. Substantial research has been conducted on carbon materials like activated carbon, carbon nanofibers, and carbon nanotubes due to their high theoretical hydrogen capacities. However, the theoretical values have not been achieved, and hydrogen uptake capacities in these materials are below 10 wt. %. In this thesis we investigated the use of silicon for hydrogen generation. Hydrogen generation via water oxidation of silicon had been ignored due to slow reaction kinetics. We hypothesized that the hydrogen generation rate could be improved by using high surface area silicon nanoparticles. Our laser-pyrolysis-produced nanoparticles showed surprisingly rapid hydrogen generation and high hydrogen yield, exceeding the theoretical maximum of two moles of H2 per

Hydrogen transmission/storage with a metal hydride/organic slurry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breault, R.W.; Rolfe, J.; McClaine, A.

1998-08-01

Thermo Power Corporation has developed a new approach for the production, transmission, and storage of hydrogen. In this approach, a chemical hydride slurry is used as the hydrogen carrier and storage media. The slurry protects the hydride from unanticipated contact with moisture in the air and makes the hydride pumpable. At the point of storage and use, a chemical hydride/water reaction is used to produce high-purity hydrogen. An essential feature of this approach is the recovery and recycle of the spent hydride at centralized processing plants, resulting in an overall low cost for hydrogen. This approach has two clear benefits:more » it greatly improves energy transmission and storage characteristics of hydrogen as a fuel, and it produces the hydrogen carrier efficiently and economically from a low cost carbon source. The preliminary economic analysis of the process indicates that hydrogen can be produced for $3.85 per million Btu based on a carbon cost of $1.42 per million Btu and a plant sized to serve a million cars per day. This compares to current costs of approximately $9.00 per million Btu to produce hydrogen from $3.00 per million Btu natural gas, and $25 per million Btu to produce hydrogen by electrolysis from $0.05 per Kwh electricity. The present standard for production of hydrogen from renewable energy is photovoltaic-electrolysis at $100 to $150 per million Btu.« less

Heat Transfer Enhancement of Metal Hydride Particle Bed for Heat Driven Type Refrigerator by Carbon Fiber

NASA Astrophysics Data System (ADS)

Bae, Sang-Chul; Tanae, Takayuki; Monde, Masanori; Katsuta, Masafumi

A series of study has been performed on the metal hydride particle beds of Ti0.15Zr0.85Cr0.9Fe0.6Ni0.2Mn0.3Cu0.05 (MH-1, using for heat source), Ti0.73Zr0.27Cr1.2Fe0.3Ni0.1Mn0.4Cu0.05 (MH-2, using for cooling load) to measure the effective thermal conductivities. The effective thermal conductivities of activated and oxidized MH particle bed in helium have been examined. Experiment results show that pressure has great influence on effective thermal conductivity in low pressure range (<0.5 MPa). And that influence decreases rapidly with increase of gas pressure. The reason of pressure dependence at low pressure range is that the mean free path of gas becomes greater than effective thickness of gas film which is important to the heat transfer mechanism of particle bed. In order to enhance the poor thermal conductivity of metal hydride particle bed, carbon fiber mixing method has been used in this study. Three types, two insert methods and five mass percentages of carbon fiber have been examined and compared. The highest effective thermal conductivity of MH particle bed has been reached with Type B carbon fiber which has second higher thermal conductivity, and 2 weight percentage. This method has acquired 5-6 times higher thermal conductivity than pure metal hydride particle beds with quite low quantity of additives, only 2 mass% of carbon fiber. This is a good result comparing to other method which can reach higher effective thermal conductivity but needs much higher percentage of additives too.

Hydrogen, lithium, and lithium hydride production

DOEpatents

Brown, Sam W.; Spencer, Larry S.; Phillips, Michael R.; Powell, G. Louis; Campbell, Peggy J.

2017-06-20

A method is provided for extracting hydrogen from lithium hydride. The method includes (a) heating lithium hydride to form liquid-phase lithium hydride; (b) extracting hydrogen from the liquid-phase lithium hydride, leaving residual liquid-phase lithium metal; (c) hydriding the residual liquid-phase lithium metal to form refined lithium hydride; and repeating steps (a) and (b) on the refined lithium hydride.

U-PuO2, U-PuC, U-PuN cermet fuel for fast reactor

NASA Astrophysics Data System (ADS)

Mishra, Sudhir; Kaity, Santu; Banerjee, Joydipta; Nandi, Chiranjeet; Dey, G. K.; Khan, K. B.

2018-02-01

Cermet fuel combines beneficial properties of both ceramic and metal and attracts global interest for research as a candidate fuel for nuclear reactors. In the present study, U matrix PuC/PuN/PuO2 cermet for fast reactor have been fabricated on laboratory scale by the powder metallurgy route. Characterization of the fuel has been carried out using Dilatometer, Differential Thermal analysis (DTA), X-ray diffractometer and Optical microscope. X ray diffraction study of the fuel reveals presence of different phases. The PuN dispersed cermet was observed to have high solidus temperature as compared to PuC and PuO2 dispersed cermet. Swelling was observed in U matrix PuO2 cermet which also showed higher thermal expansion. Among the three cermets studied, U matrix PuC cermet showed maximum thermal conductivity.

Properties of nanostructured undoped ZrO{sub 2} thin film electrolytes by plasma enhanced atomic layer deposition for thin film solid oxide fuel cells

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cho, Gu Young; Noh, Seungtak; Lee, Yoon Ho

2016-01-15

Nanostructured ZrO{sub 2} thin films were prepared by thermal atomic layer deposition (ALD) and by plasma-enhanced atomic layer deposition (PEALD). The effects of the deposition conditions of temperature, reactant, plasma power, and duration upon the physical and chemical properties of ZrO{sub 2} films were investigated. The ZrO{sub 2} films by PEALD were polycrystalline and had low contamination, rough surfaces, and relatively large grains. Increasing the plasma power and duration led to a clear polycrystalline structure with relatively large grains due to the additional energy imparted by the plasma. After characterization, the films were incorporated as electrolytes in thin film solidmore » oxide fuel cells, and the performance was measured at 500 °C. Despite similar structure and cathode morphology of the cells studied, the thin film solid oxide fuel cell with the ZrO{sub 2} thin film electrolyte by the thermal ALD at 250 °C exhibited the highest power density (38 mW/cm{sup 2}) because of the lowest average grain size at cathode/electrolyte interface.« less

Study of the mechanical behavior of the hydride blister/rim structure in Zircaloy-4 using in-situ synchrotron X-ray diffraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Jun-li; Han, Xiaochun; Heuser, Brent J.

2016-04-01

High-energy synchrotron X-ray diffraction was utilized to study the mechanical response of the f.c.c delta hydride phase, the intermetallic precipitation with hexagonal C14 lave phase and the alpha-Zr phase in the Zircaloy-4 materials with a hydride rim/blister structure near one surface of the material during in-situ uniaxial tension experiment at 200 degrees C. The f.c.c delta was the only hydride phase observed in the rim/blister structure. The conventional Rietveld refinement was applied to measure the macro-strain equivalent response of the three phases. Two regions were delineated in the applied load versus lattice strain measurement: a linear elastic strain region andmore » region that exhibited load partitioning. Load partitioning was quantified by von Mises analysis. The three phases were observed to have similar elastic modulus at 200 degrees C.« less

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

Oxidation kinetics of hydride-bearing uranium metal corrosion products

NASA Astrophysics Data System (ADS)

Totemeier, Terry C.; Pahl, Robert G.; Frank, Steven M.

The oxidation behavior of hydride-bearing uranium metal corrosion products from Zero Power Physics Reactor (ZPPR) fuel plates was studied using thermo-gravimetric analysis (TGA) in environments of Ar-4%O 2, Ar-9%O 2, and Ar-20%O 2. Ignition of corrosion product samples from two moderately corroded plates was observed between 125°C and 150°C in all environments. The rate of oxidation above the ignition temperature was found to be dependent only on the net flow rate of oxygen in the reacting gas. Due to the higher net oxygen flow rate, burning rates increased with increasing oxygen concentration. Oxidation rates below the ignition temperature were much slower and decreased with increasing test time. The hydride contents of the TGA samples from the two moderately corroded plates, determined from the total weight gain achieved during burning, were 47-61 wt% and 29-39 wt%. Samples from a lightly corroded plate were not reactive; X-ray diffraction (XRD) confirmed that they contained little hydride.

Production of ZrC Matrix for Use in Gas Fast Reactor Composite Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vasudevamurthy, Gokul; Knight, Travis W.; Roberts, Elwyn

2007-07-01

Zirconium carbide is being considered as a candidate for inert matrix material in composite nuclear fuel for Gas fast reactors due to its favorable characteristics. ZrC can be produced by the direct reaction of pure zirconium and graphite powders. Such a reaction is exothermic in nature. The reaction is self sustaining once initial ignition has been achieved. The heat released during the reaction is high enough to complete the reaction and achieve partial sintering without any external pressure applied. External heat source is required to achieve ignition of the reactants and maintain the temperature close to the adiabatic temperature tomore » achieve higher levels of sintering. External pressure is also a driving force for sintering. In the experiments described, cylindrical compacts of ZrC were produced by direct combustion reaction. External induction heating combined with varying amounts of external applied pressure was employed to achieve varying degrees of density/porosity. The effect of reactant particle size on the product characteristics was also studied. The samples were characterized for density/porosity, composition and microstructure. (authors)« less

Boron hydride polymer coated substrates

DOEpatents

Pearson, R.K.; Bystroff, R.I.; Miller, D.E.

1986-08-27

A method is disclosed for coating a substrate with a uniformly smooth layer of a boron hydride polymer. The method comprises providing a reaction chamber which contains the substrate and the boron hydride plasma. A boron hydride feed stock is introduced into the chamber simultaneously with the generation of a plasma discharge within the chamber. A boron hydride plasma of ions, electrons and free radicals which is generated by the plasma discharge interacts to form a uniformly smooth boron hydride polymer which is deposited on the substrate.

Boron hydride polymer coated substrates

DOEpatents

Pearson, Richard K.; Bystroff, Roman I.; Miller, Dale E.

1987-01-01

A method is disclosed for coating a substrate with a uniformly smooth layer of a boron hydride polymer. The method comprises providing a reaction chamber which contains the substrate and the boron hydride plasma. A boron hydride feed stock is introduced into the chamber simultaneously with the generation of a plasma discharge within the chamber. A boron hydride plasma of ions, electrons and free radicals which is generated by the plasma discharge interacts to form a uniformly smooth boron hydride polymer which is deposited on the substrate.

Method of production of pure hydrogen near room temperature from aluminum-based hydride materials

DOEpatents

Pecharsky, Vitalij K.; Balema, Viktor P.

2004-08-10

The present invention provides a cost-effective method of producing pure hydrogen gas from hydride-based solid materials. The hydride-based solid material is mechanically processed in the presence of a catalyst to obtain pure gaseous hydrogen. Unlike previous methods, hydrogen may be obtained from the solid material without heating, and without the addition of a solvent during processing. The described method of hydrogen production is useful for energy conversion and production technologies that consume pure gaseous hydrogen as a fuel.

17. VIEW OF HYDRIDING SYSTEM IN BUILDING 881. THE HYDRIDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

17. VIEW OF HYDRIDING SYSTEM IN BUILDING 881. THE HYDRIDING SYSTEM WAS PART OF THE FAST ENRICHED URANIUM RECOVERY PROCESS. (11/11/59) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

New Possibilities for Understanding Complex Metal Hydrides via Synchrotron X-ray Studies

NASA Astrophysics Data System (ADS)

Dobbins, Tabbetha

2008-03-01

Ultrasmall-angle x-ray scattering (USAXS) and X-ray absorption spectroscopy (XAS) are used for the study of chemical and morphological changes in metal hydride powder (e.g. NaAlH4) both before and after transition metal salt catalytic dopant additions by high energy ball milling. The variation in surface fractal dimension and particle size with milling time and dopant content were tracked. These studies show that dopant content level (e.g. 2 mol % and 4 mol %) and dopant type (i.e. TiCl2, TiCl3, VCl3, and ZrCl4) markedly affects NaAlH4 powder particle surface area (determined using USAXS surface fractal dimension). As well, the chemical reaction between the catalyst and hydride powder was further elucidated using XAS. Ti-metal reacts with the Al desorption product (from NaAlH4) to form TiAlx product phases. These studies were able to link powder particle surface area to catalytic doping and were able to link dopant chemical state with dehydrogenation reactant and product phases.

Nanocomposite membranes based on polybenzimidazole and ZrO2 for high-temperature proton exchange membrane fuel cells.

PubMed

Nawn, Graeme; Pace, Giuseppe; Lavina, Sandra; Vezzù, Keti; Negro, Enrico; Bertasi, Federico; Polizzi, Stefano; Di Noto, Vito

2015-04-24

Owing to the numerous benefits obtained when operating proton exchange membrane fuel cells at elevated temperature (>100 °C), the development of thermally stable proton exchange membranes that demonstrate conductivity under anhydrous conditions remains a significant goal for fuel cell technology. This paper presents composite membranes consisting of poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI4N) impregnated with a ZrO2 nanofiller of varying content (ranging from 0 to 22 wt %). The structure-property relationships of the acid-doped and undoped composite membranes have been studied using thermogravimetric analysis, differential scanning calorimetry, dynamic mechanical analysis, wide-angle X-ray scattering, infrared spectroscopy, and broadband electrical spectroscopy. Results indicate that the level of nanofiller has a significant effect on the membrane properties. From 0 to 8 wt %, the acid uptake as well as the thermal and mechanical properties of the membrane increase. As the nanofiller level is increased from 8 to 22 wt % the opposite effect is observed. At 185 °C, the ionic conductivity of [PBI4N(ZrO2 )0.231 ](H3 PO4 )13 is found to be 1.04×10(-1)  S cm(-1) . This renders membranes of this type promising candidates for use in high-temperature proton exchange membrane fuel cells. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Investigation of Lithium Metal Hydride Materials for Mitigation of Deep Space Radiation

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Atwell, William

2016-01-01

Radiation exposure to crew, electronics, and non-metallic materials is one of many concerns with long-term, deep space travel. Mitigating this exposure is approached via a multi-faceted methodology focusing on multi-functional materials, vehicle configuration, and operational or mission constraints. In this set of research, we are focusing on new multi-functional materials that may have advantages over traditional shielding materials, such as polyethylene. Metal hydride materials are of particular interest for deep space radiation shielding due to their ability to store hydrogen, a low-Z material known to be an excellent radiation mitigator and a potential fuel source. We have previously investigated 41 different metal hydrides for their radiation mitigation potential. Of these metal hydrides, we found a set of lithium hydrides to be of particular interest due to their excellent shielding of galactic cosmic radiation. Given these results, we will continue our investigation of lithium hydrides by expanding our data set to include dose equivalent and to further understand why these materials outperformed polyethylene in a heavy ion environment. For this study, we used HZETRN 2010, a one-dimensional transport code developed by NASA Langley Research Center, to simulate radiation transport through the lithium hydrides. We focused on the 1977 solar minimum Galactic Cosmic Radiation environment and thicknesses of 1, 5, 10, 20, 30, 50, and 100 g/cm2 to stay consistent with our previous studies. The details of this work and the subsequent results will be discussed in this paper.

Theoretical study of the promotional effect of ZrO2 on In2O3 catalyzed methanol synthesis from CO2 hydrogenation

NASA Astrophysics Data System (ADS)

Zhang, Minhua; Dou, Maobin; Yu, Yingzhe

2018-03-01

Methanol synthesis from CO2 hydrogenation on the ZrO2 doped In2O3(110) surface (Zr-In2O3(110)) with oxygen vacancy has been studied using the density functional theory calculations. The calculated results show that the doped ZrO2 species prohibits the excessive formation of oxygen vacancies and dissociation of H2 on In2O3 surface slightly, but enhances the adsorption of CO2 on both perfect and defective Zr-In2O3(110) surface. Methanol is formed via the HCOO route. The hydrogenation of CO2 to HCOO is both energetically and kinetically facile. The HCOO hydrogenates to polydentate H2CO (p-H2CO) species with an activation barrier of 0.75 eV. H3CO is produced from the hydrogenation of monodentate H2CO (mono-H2CO), transformation from p-H2CO with 0.82 eV reaction energy, with no barrier whether there is hydroxyl group between the mono-H2CO and the neighboring hydride or not. Methanol is the product of H3CO protonation with 0.75 eV barrier. The dissociation and protonation of CO2 are both energetically and kinetically prohibited on Zr-In2O3(110) surface. The doped ZrO2 species can further enhance the adsorption of all the intermediates involved in CO2 hydrogenation to methanol, activate the adsorbed CO2 and H2CO, and stabilize the HCOO, H2CO and H3CO, especially prohibit the dissociation of H2CO or the reaction of H2CO with neighboring hydride to form HCOO and gas phase H2. All these effects make the ZrO2 supported In2O3 catalyst exhibit higher activity and selectivity on methanol synthesis from CO2 hydrogenation.

[U.S. renewable fuel standard implementation mechanism and market tracking].

PubMed

Kang, Liping; Earley, Robert; An, Feng; Zhang, Yu

2013-03-01

U.S. Renewable Fuel Standard (RFS) is a mandatory policy for promoting the utilization of biofuels in road transpiration sector in order to reduce the country's dependency on foreign oil and greenhouse gas emissions. U.S. Environmental Protection Agency (EPA) defines the proportion of renewable fuels according to RFS annual target, and requests obligated parties such like fossil fuel refiner, blenders and importer in the U.S. to complete Renewable Volume Obligation (RVO) every year. Obligated parties prove they have achieved their RVO through a renewable fuels certification system, which generates Renewable Identification Numbers (RINs) for every gallon of qualified renewable fuels produced or imported into U.S., RINs is a key for tracking renewable fuel consumption, which in turn is a key for implementing the RFS in the U.S., separated RINs can be freely traded in market and obligated parties could fulfill their RVO through buying RINs from other stakeholders. This briefing paper highlights RFS policy implementing mechanism and marketing tracking, mainly describes importance of RINs, and the method for generating and tracking RINs by both government and fuels industry participants.

In-Bed Accountability Development for a Passively Cooled, Electrically Heated Hydride (PACE) Bed

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, J.E.

A nominal 1500 STP-L PAssively Cooled, Electrically heated hydride (PACE) Bed has been developed for implementation into a new Savannah River Site tritium project. The 1.2 meter (four-foot) long process vessel contains on internal 'U-tube' for tritium In-Bed Accountability (IBA) measurements. IBA will be performed on six, 12.6 kg production metal hydride storage beds.IBA tests were done on a prototype bed using electric heaters to simulate the radiolytic decay of tritium. Tests had gas flows from 10 to 100 SLPM through the U-tube or 100 SLPM through the bed's vacuum jacket. IBA inventory measurement errors at the 95% confidence levelmore » were calculated using the correlation of IBA gas temperature rise, or (hydride) bed temperature rise above ambient temperature, versus simulated tritium inventory.Prototype bed IBA inventory errors at 100 SLPM were the largest for gas flows through the vacuum jacket: 15.2 grams for the bed temperature rise and 11.5 grams for the gas temperature rise. For a 100 SLPM U-tube flow, the inventory error was 2.5 grams using bed temperature rise and 1.6 grams using gas temperature rise. For 50 to 100 SLPM U-tube flows, the IBA gas temperature rise inventory errors were nominally one to two grams that increased above four grams for flows less than 50 SLPM. For 50 to 100 SLPM U-tube flows, the IBA bed temperature rise inventory errors were greater than the gas temperature rise errors, but similar errors were found for both methods at gas flows of 20, 30, and 40 SLPM.Electric heater IBA tests were done for six production hydride beds using a 45 SLPM U-tube gas flow. Of the duplicate runs performed on these beds, five of the six beds produced IBA inventory errors of approximately three grams: consistent with results obtained in the laboratory prototype tests.« less

In-Bed Accountability Development for a Passively Cooled, Electrically Heated Hydride (PACE) Bed

DOE Office of Scientific and Technical Information (OSTI.GOV)

KLEIN, JAMES

A nominal 1500 STP-L PAssively Cooled, Electrically heated hydride (PACE) Bed has been developed for implementation into a new Savannah River Site tritium project. The 1.2 meter (four-foot) long process vessel contains an internal ''U-tube'' for tritium In-Bed Accountability (IBA) measurements. IBA will be performed on six, 12.6 kg production metal hydride storage beds. IBA tests were done on a prototype bed using electric heaters to simulate the radiolytic decay of tritium. Tests had gas flows from 10 to 100 SLPM through the U-tube or 100 SLPM through the bed's vacuum jacket. IBA inventory measurement errors at the 95 percentmore » confidence level were calculated using the correlation of IBA gas temperature rise, or (hydride) bed temperature rise above ambient temperature, versus simulated tritium inventory. Prototype bed IBA inventory errors at 100 SLPM were the largest for gas flows through the vacuum jacket: 15.2 grams for the bed temperature rise and 11.5 grams for the gas temperature rise. For a 100 SLPM U-tube flow, the inventory error was 2.5 grams using bed temperature rise and 1.6 grams using gas temperature rise. For 50 to 100 SLPM U-tube flows, the IBA gas temperature rise inventory errors were nominally one to two grams that increased above four grams for flows less than 50 SLPM. For 50 to 100 SLPM U-tube flows, the IBA bed temperature rise inventory errors were greater than the gas temperature rise errors, but similar errors were found for both methods at gas flows of 20, 30, and 40 SLPM. Electric heater IBA tests were done for six production hydride beds using a 45 SLPM U-tube gas flow. Of the duplicate runs performed on these beds, five of the six beds produced IBA inventory errors of approximately three grams: consistent with results obtained in the laboratory prototype tests.« less

Heterogeneous reduction of carbon dioxide by hydride-terminated silicon nanocrystals

PubMed Central

Sun, Wei; Qian, Chenxi; He, Le; Ghuman, Kulbir Kaur; Wong, Annabelle P. Y.; Jia, Jia; Jelle, Abdinoor A.; O'Brien, Paul G.; Reyes, Laura M.; Wood, Thomas E.; Helmy, Amr S.; Mims, Charles A.; Singh, Chandra Veer; Ozin, Geoffrey A.

2016-01-01

Silicon constitutes 28% of the earth's mass. Its high abundance, lack of toxicity and low cost coupled with its electrical and optical properties, make silicon unique among the semiconductors for converting sunlight into electricity. In the quest for semiconductors that can make chemicals and fuels from sunlight and carbon dioxide, unfortunately the best performers are invariably made from rare and expensive elements. Here we report the observation that hydride-terminated silicon nanocrystals with average diameter 3.5 nm, denoted ncSi:H, can function as a single component heterogeneous reducing agent for converting gaseous carbon dioxide selectively to carbon monoxide, at a rate of hundreds of μmol h−1 g−1. The large surface area, broadband visible to near infrared light harvesting and reducing power of SiH surface sites of ncSi:H, together play key roles in this conversion. Making use of the reducing power of nanostructured hydrides towards gaseous carbon dioxide is a conceptually distinct and commercially interesting strategy for making fuels directly from sunlight. PMID:27550234

Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

2011-12-01

The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

CERDEC Fuel Cell Team: Military Transitions for Soldier Fuel Cells

DTIC Science & Technology

2008-10-27

Fuel Cell (DMFC) (PEO Soldier) Samsung: 20W DMFC (CRADA) General Atomics & Jadoo: 50W Ammonia Borane Fueled PEMFC Current Fuel Cell Team Efforts...Continued Ardica: 20W Wearable PEMFC operating on Chemical Hydrides Spectrum Brands w/ Rayovac: Hydrogen Generators and Alkaline Fuel Cells for AA...100W Ammonia Borane fueled PEMFC Ultralife: 150W sodium borohydride fueled PEMFC Protonex: 250W RMFC and Power Manager (ARO) NanoDynamics: 250W SOFC

Fourier Transform IR Spectroscopic Study of Nano-ZrO2 + Nano-SiO2 + Nano-H2O Systems Upon the Action of Gamma Radiation

NASA Astrophysics Data System (ADS)

Agayev, T. N.; Gadzhieva, N. N.; Melikova, S. Z.

2018-05-01

The radiation decomposition of water in a nano-ZrO2 + nano-SiO2 + H2O system at 300 K by the action of gamma radiation has been studied by Fourier transform IR spectroscopy. Water adsorption in the zirconium and silicon nanooxides is attributed to molecular and dissociative mechanisms. Active intermediates in this radiation-induced heterogeneous decomposition of water were detected including zirconium and silicon hydrides and hydroxyl groups. Variation in the ratio of ZrO2 and SiO2 nanopowders was shown to lead to change in their radiation catalytic activity compared to initial ZrO2.

Thermodynamic Hydricity of Transition Metal Hydrides

DOE PAGES

Wiedner, Eric S.; Chambers, Matthew B.; Pitman, Catherine L.; ...

2016-08-02

Transition metal hydrides play a critical role in stoichiometric and catalytic transformations. Knowledge of free energies for cleaving metal hydride bonds enables the prediction of chemical reactivity, such as for the bond-forming and bondbreaking events that occur in a catalytic reaction. Thermodynamic hydricity is the free energy required to cleave an M-H bond to generate a hydride ion (H -). Three primary methods have been developed for hydricity determination: the hydride transfer method establishes hydride transfer equilibrium with a hydride donor/acceptor pair of known hydricity, the H 2 heterolysis method involves measuring the equilibrium of heterolytic cleavage of H 2more » in the presence of a base, and the potential-pK a method considers stepwise transfer of a proton and two electrons to give a net hydride transfer. Using these methods, over 100 thermodynamic hydricity values for transition metal hydrides have been determined in acetonitrile or water. In acetonitrile, the hydricity of metal hydrides spans a range of more than 50 kcal/mol. Finally, methods for using hydricity values to predict chemical reactivity are also discussed, including organic transformations, the reduction of CO 2, and the production and oxidation of hydrogen.« less

High-temperature deformation and processing maps of Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles

NASA Astrophysics Data System (ADS)

Chen, Jing; Liu, Huiqun; Zhang, Ruiqian; Li, Gang; Yi, Danqing; Lin, Gaoyong; Guo, Zhen; Liu, Shaoqiang

2018-06-01

High-temperature compression deformation of a Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles was investigated at 750 °C-950 °C with a strain rate of 0.01-1.0 s-1 and height reduction of 20%. Scanning electron microscopy was utilized to investigate the influence of the deformation conditions on the microstructure of the composite and damage to the coated surrogate fuel particles. The results indicated that the flow stress of the composite increased with increasing strain rate and decreasing temperature. The true stress-strain curves showed obvious serrated oscillation characteristics. There were stable deformation ranges at the initial deformation stage with low true strain at strain rate 0.01 s-1 for all measured temperatures. Additionally, the coating on the surface of the surrogate nuclear fuel particles was damaged when the Zr-4 matrix was deformed at conditions of high strain rate and low temperature. The deformation stability was obtained from the processing maps and microstructural characterization. The high-temperature deformation activation energy was 354.22, 407.68, and 433.81 kJ/mol at true strains of 0.02, 0.08, and 0.15, respectively. The optimum deformation parameters for the composite were 900-950 °C and 0.01 s-1. These results are expected to provide guidance for subsequent determination of possible hot working processes for this composite.

Low-pressure Structural Modification of Aluminum Hydride

DTIC Science & Technology

2011-02-01

Acknowledgments Use of the National Synchrotron Light Source (NSLS), Brookhaven National Laboratory ( BNL ) was supported by the U.S. Department of Energy...National Synchrotron Light Source (NSLS) of Brookhaven National Laboratory ( BNL ). The spectral resolution of ±4 cm–1 was used for all IR measurements...12 List of Symbols, Abbreviations, and Acronyms Al aluminum AlH3 aluminum hydride BNL Brookhaven National Laboratory EOS equation of

Crack growth through the thickness of thin-sheet Hydrided Zircaloy-4

NASA Astrophysics Data System (ADS)

Raynaud, Patrick A. C.

In recent years, the limits on fuel burnup have been increased to allow an increase in the amount of energy produced by a nuclear fuel assembly thus reducing waste volume and allowing greater capacity factors. As a result, it is paramount to ensure safety after longer reactor exposure times in the case of design-basis accidents, such as reactivity-initiated accidents (RIA). Previously proposed failure criteria do not directly address the particular cladding failure mechanism during a RIA, in which crack initiation in brittle outer-layers is immediately followed by crack growth through the thickness of the thin-wall tubing. In such a case, the fracture toughness of hydrided thin-wall cladding material must be known for the conditions of through-thickness crack growth in order to predict the failure of high-burnup cladding. The fracture toughness of hydrided Zircaloy-4 in the form of thin-sheet has been examined for the condition of through-thickness crack growth as a function of hydride content and distribution at 25°C, 300°C, and 375°C. To achieve this goal, an experimental procedure was developed in which a linear hydride blister formed across the width of a four-point bend specimen was used to inject a sharp crack that was subsequently extended by fatigue pre-cracking. The electrical potential drop method was used to monitor the crack length during fracture toughness testing, thus allowing for correlation of the load-displacement record with the crack length. Elastic-plastic fracture mechanics were used to interpret the experimental test results in terms of fracture toughness, and J-R crack growth resistance curves were generated. Finite element modeling was performed to adapt the classic theories of fracture mechanics applicable to thick-plate specimens to the case of through-thickness crack growth in thin-sheet materials, and to account for non-uniform crack fronts. Finally, the hydride microstructure was characterized in the vicinity of the crack tip by

Method for preparing porous metal hydride compacts

DOEpatents

Ron, Moshe; Gruen, Dieter M.; Mendelsohn, Marshall H.; Sheft, Irving

1981-01-01

A method for preparing porous metallic-matrix hydride compacts which can be repeatedly hydrided and dehydrided without disintegration. A mixture of a finely divided metal hydride and a finely divided matrix metal is contacted with a poison which prevents the metal hydride from dehydriding at room temperature and atmospheric pressure. The mixture of matrix metal and poisoned metal hydride is then compacted under pressure at room temperature to form porous metallic-matrix hydride compacts.

Method for preparing porous metal hydride compacts

DOEpatents

Ron, M.; Gruen, D.M.; Mendelsohn, M.H.; Sheft, I.

1980-01-21

A method for preparing porous metallic-matrix hydride compacts which can be repeatedly hydrided and dehydrided without disintegration. A mixture of a finely divided metal hydride and a finely divided matrix metal is contacted with a poison which prevents the metal hydride from dehydriding at room temperature and atmospheric pressure. The mixture of matrix metal and poisoned metal hydride is then compacted under pressure at room temperature to form porous metallic-matrix hydride compacts.

Influence of free carbon on the characteristics of ZrC and deposition of near-stoichiometric ZrC in TRISO coated particle fuel

NASA Astrophysics Data System (ADS)

Kim, Daejong; Ko, Myeong Jin; Park, Ji Yeon; Cho, Moon Sung; Kim, Weon-Ju

2014-08-01

Advanced TRISO coated particles with a ZrC coating layer as a main pressure boundary were fabricated by a fluidized-bed chemical vapor deposition (FBCVD) method using a chloride process. Experiments were performed to determine the effect of codeposition of graphitic carbon on the hardness and obtain the stoichiometric ZrC phase. The ZrC coating layer was composed of a mixture of ZrC and graphitic carbon phases at a low ZrCl4/CH4 ratio. A near-stoichiometric ZrC without the free carbon can be obtained by employing an impeller-driven ZrCl4 vaporizer. The codeposition of the graphitic carbon significantly lowered the hardness of ZrC while increasing the fraction of the carbon. The hardness reached its maximum when ZrC was in a slight carbon deficit without free carbon. As the graphitic carbon increased up to 12 vol%, the hardness was reduced by approximately 50% compared to the near-stoichiometric ZrC.

Analysis of Transportation Options for Commercial Spent Fuel in the U.S.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena; Busch, Ingrid Karin

.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and

Pore growth in U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D.-S.; Jamison, L. M.

2016-09-01

U-Mo/Al dispersion fuel is currently under development in the DOE's Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

Durability test on irradiated rock-like oxide fuels

NASA Astrophysics Data System (ADS)

Kuramoto, K.; Nitani, N.; Yamashita, T.

2003-06-01

For a profitable use of Pu, Japan Atomic Energy Research Institute has been promoting researches for once-through type fuels. The strategy consists of stable rock-like oxide fuel fabrication in conventional fuel facilities followed by almost complete Pu burning in LWR and disposal of chemically stable spent fuel without further processing. Because leach rates of hazardous nuclides, such as TRU and β-emitters, that have long half-lives, are very important for the evaluation of geological safety, leaching tests in deionized water at 363 K were performed with reference to the MCC-1 method. Five irradiated fuel pellets, a single phase fuel of a yttria-stabilized zirconia (YSZ) containing UO 2 (U-YSZ), two fuels of U-YSZ particle dispersed in MgAl 2O 4 (SPI) or Al 2O 3 (COR) matrix, two homogeneous-blended fuels of U-YSZ and SPI or COR powders, were submitted to the tests. Stainless steel containers with Au coating and ethylene propylene diene monomer were used as leaching vessels and packing, respectively. The evaluated normalized leach rates of Zr, U and Pu were obviously lower than those of the other important elements and nuclides. Americium, Np and especially Y showed unexpectedly high evaluated normalized leach rates. The volatile elements, Cs and I, showed enhanced leaching within particle-dispersed type fuels because of crack formation around the particle.

Structure and functional properties of TiNiZr surface layers obtained by high-velocity oxygen fuel spraying

NASA Astrophysics Data System (ADS)

Rusinov, P. O.; Blednova, Zh M.; Borovets, O. I.

2017-05-01

The authors studied a complex method of surface modification of steels for materials with shape memory effect (SME) Ti-Ni-Zr with a high-velocity oxygen-fuel spraying (HVOF) of mechanically activated (MA) powder in a protective medium. We assessed the functional properties and X-ray diffraction studies, which showed that the formation of surface layers according to the developed technology ensures the manifestation of the shape memory effect.

SOLID SOLUTION CARBIDES ARE THE KEY FUELS FOR FUTURE NUCLEAR THERMAL PROPULSION

NASA Technical Reports Server (NTRS)

Panda, Binayak; Hickman, Robert R.; Shah, Sandeep

2005-01-01

Nuclear thermal propulsion uses nuclear energy to directly heat a propellant (such as liquid hydrogen) to generate thrust for space transportation. In the 1960 s, the early Rover/Nuclear Engine for Rocket Propulsion Application (NERVA) program showed very encouraging test results for space nuclear propulsion but, in recent years, fuel research has been dismal. With NASA s renewed interest in long-term space exploration, fuel researchers are now revisiting the RoverMERVA findings, which indicated several problems with such fuels (such as erosion, chemical reaction of the fuel with propellant, fuel cracking, and cladding issues) that must be addressed. It is also well known that the higher the temperature reached by a propellant, the larger the thrust generated from the same weight of propellant. Better use of fuel and propellant requires development of fuels capable of reaching very high temperatures. Carbides have the highest melting points of any known material. Efforts are underway to develop carbide mixtures and solid solutions that contain uranium carbide, in order to achieve very high fuel temperatures. Binary solid solution carbides (U, Zr)C have proven to be very effective in this regard. Ternary carbides such as (U, Zr, X) carbides (where X represents Nb, Ta, W, and Hf) also hold great promise as fuel material, since the carbide mixtures in solid solution generate a very hard and tough compact material. This paper highlights past experience with early fuel materials and bi-carbides, technical problems associated with consolidation of the ingredients, and current techniques being developed to consolidate ternary carbides as fuel materials.

Development of metal hydride composites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, J.W.

1992-12-01

Most of current hydride technology at Savannah River Site is based on beds of metal hydride powders; the expansion upon hydridation and the cycling results in continued breakdown into finer particles. Goal is to develop a composite which will contain the fines in a dimensionally stable matrix, for use in processes which require a stable gas flow through a hydride bed. Metal hydride composites would benefit the advanced Thermal Cycling Absorption process (hydrogen isotope separation), and the Replacement Tritium Facility (storage, pumping, compression, purification of hydrogen isotopes). These composites were fabricated by cold compaction of a mixture of metal hydridemore » granules and coarse copper powder; the porosity in the granules was introduced by means of ammonium carbonate. The composite pellets were cycled 138 times in hydrogen with the loss of LANA0.75 (LaNi{sub 4.25}Al{sub 0.75}) limited to the surface. Vacuum sintering can provide additional strength at the edges. Without a coating, the metal hydride particles exposed at the pellet surface can be removed by cycling several times in hydrogen.« less

System for operating solid oxide fuel cell generator on diesel fuel

NASA Technical Reports Server (NTRS)

Singh, Prabhu (Inventor); George, Raymond A. (Inventor)

1997-01-01

A system is provided for operating a solid oxide fuel cell generator on diesel fuel. The system includes a hydrodesulfurizer which reduces the sulfur content of commercial and military grade diesel fuel to an acceptable level. Hydrogen which has been previously separated from the process stream is mixed with diesel fuel at low pressure. The diesel/hydrogen mixture is then pressurized and introduced into the hydrodesulfurizer. The hydrodesulfurizer comprises a metal oxide such as ZnO which reacts with hydrogen sulfide in the presence of a metal catalyst to form a metal sulfide and water. After desulfurization, the diesel fuel is reformed and delivered to a hydrogen separator which removes most of the hydrogen from the reformed fuel prior to introduction into a solid oxide fuel cell generator. The separated hydrogen is then selectively delivered to the diesel/hydrogen mixer or to a hydrogen storage unit. The hydrogen storage unit preferably comprises a metal hydride which stores hydrogen in solid form at low pressure. Hydrogen may be discharged from the metal hydride to the diesel/hydrogen mixture at low pressure upon demand, particularly during start-up and shut-down of the system.

Impedance and self-discharge mechanism studies of nickel metal hydride batteries for energy storage applications

NASA Astrophysics Data System (ADS)

Zhu, Wenhua; Zhu, Ying; Tatarchuk, Bruce

2013-04-01

Nickel metal hydride battery packs have been found wide applications in the HEVs (hybrid electric vehicles) through the on-board rapid energy conservation and efficient storage to decrease the fossil fuel consumption rate and reduce CO2 emissions as well as other harmful exhaust gases. In comparison to the conventional Ni-Cd battery, the Ni-MH battery exhibits a relatively higher self-discharge rate. In general, there are quite a few factors that speed up the self-discharge of the electrodes in the sealed nickel metal hydride batteries. This disadvantage eventually reduces the overall efficiency of the energy conversion and storage system. In this work, ac impedance data were collected from the nickel metal hydride batteries. The self-discharge mechanism and battery capacity degradation were analyzed and discussed for further performance improvement.

Am phases in the matrix of a U–Pu–Zr alloy with Np, Am, and rare-earth elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn E.; Kennedy, J. Rory; Madden, James W.

2015-01-01

Phases and microstructures in the matrix of an as-cast U-Pu-Zr alloy with 3 wt% Am, 2% Np, and 8% rare-earth elements were characterized by scanning and transmission electron microscopy. The matrix consists primarily of two phases, both of which contain Am: ζ-(U, Np, Pu, Am) (~70 at% U, 5% Np, 14% Pu, 1% Am, and 10% Zr) and δ-(U, Np, Pu, Am)Zr 2 (~25% U, 2% Np, 10-15% Pu, 1-2% Am, and 55-60 at% Zr). These phases are similar to those in U-Pu-Zr alloys, although the Zr content in ζ-(U, Np, Pu, Am) is higher than that in ζ-(U, Pu)more » and the Zr content in δ-(U, Np, Pu, Am)Zr 2 is lower than that in δ-UZr 2. Nanocrystalline actinide oxides with structures similar to UO2 occurred in some areas, but may have formed by reactions with the atmosphere during sample handling. Planar features consisting of a central zone of ζ-(U, Np, Pu, Am) bracketed by zones of δ-(U, Np, Pu, Am)Zr 2 bound irregular polygons ranging in size from a few micrometers to a few tens of micrometers across. The rest of the matrix consists of elongated domains of ζ-(U, Np, Pu, Am) and δ-(U, Np, Pu, Am)Zr 2. Each of these domains is a few tens of nanometers across and a few hundred nanometers long. The domains display strong preferred orientations involving areas a few hundred nanometers to a few micrometers across.« less

Materials for Hydrogen Storage: From Nanostructures to Complex Hydrides

NASA Astrophysics Data System (ADS)

Jena, Puru

2006-03-01

The limited supply of fossil fuels, its adverse effect on the environment, and growing worldwide demand for energy has necessitated the search for new and clean sources of energy. The possibility of using hydrogen to meet this growing energy need has rekindled interest in the study of safe, efficient, and economical storage of hydrogen. This talk will discuss the issues and challenges in storing hydrogen in light complex hydrides and discuss the role of nanostructuring and catalysts that can improve the thermodynamics and kinetics of hydrogen. In particular, we will discuss how studies of clusters can help elucidate the fundamental mechanisms for hydrogen storage and how these can be applied in Boron Nitride and Carbon nanocages and how metallization of these nanostructures is necessary to store hydrogen with large gravimetric density. We will also discuss the properties of complex light metal hydrides such as alanates and magnesium hydrides that can store up to 18 wt % hydrogen, although the temperature where hydrogen desorbs is rather high. Using first principles calculations, we will provide a fundamental understanding of the electronic structure and stability of these systems and how it is affected due to catalysts. It is hoped that the understanding gained here can be useful in designing better catalysts as well as hosts for hydrogen storage.

Vanadium hydride deuterium-tritium generator

DOEpatents

Christensen, Leslie D.

1982-01-01

A pressure controlled vanadium hydride gas generator to provide deuterium-tritium gas in a series of pressure increments. A high pressure chamber filled with vanadium-deuterium-tritium hydride is surrounded by a heater which controls the hydride temperature. The heater is actuated by a power controller which responds to the difference signal between the actual pressure signal and a programmed pressure signal.

Hydrogen /Hydride/-air secondary battery

NASA Technical Reports Server (NTRS)

Sarradin, J.; Bronoel, G.; Percheron-Guegan, A.; Achard, J. C.

1979-01-01

The use of metal hydrides as negative electrodes in a hydrogen-air secondary battery seems promising. However, in an unpressurized cell, more stable hydrides that LaNi5H6 must be selected. Partial substitutions of nickel by aluminium or manganese increase the stability of hydrides. Combined with an air reversible electrode, a specific energy close to 100 Wh/kg can be expected.

Vanadium hydride deuterium-tritium generator

DOEpatents

Christensen, L.D.

1980-03-13

A pressure controlled vanadium hydride gas generator was designed to provide deuterium-tritium gas in a series of pressure increments. A high pressure chamber filled with vanadium-deuterium-tritium hydride is surrounded by a heater which controls the hydride temperature. The heater is actuated by a power controller which responds to the difference signal between the actual pressure signal and a programmed pressure signal.

Hydrogen, lithium, and lithium hydride production

DOEpatents

Brown, Sam W; Spencer, Larry S; Phillips, Michael R; Powell, G. Louis; Campbell, Peggy J

2014-03-25

A method of producing high purity lithium metal is provided, where gaseous-phase lithium metal is extracted from lithium hydride and condensed to form solid high purity lithium metal. The high purity lithium metal may be hydrided to provide high purity lithium hydride.

Fluorite transition metal hydride induced destabilization of the MgH2 system in MgH2/TMH2 multilayers ( TM=Sc , Ti, V, Cr, Y, Zr, Nb, La, Hf)

NASA Astrophysics Data System (ADS)

Tao, S. X.; Notten, P. H. L.; van Santen, R. A.; Jansen, A. P. J.

2010-09-01

The structural changes in MgH2 induced by contact with fluorite transition metal hydrides ( TMH2 , TM=Sc , Ti, V, Cr, Y, Zr, Nb, La, Hf) have been studied using density-functional theory calculations. Models of MgH2(rutile)/TiH2(fluorite) and MgH2(fluorite)/TiH2(fluorite) multilayers with different Mg:TM ratios have been designed. With a fixed thickness of the TMH2 layer, structure transformation of MgH2 from rutile to fluorite occurs with a decrease in thickness of the MgH2 layer. The hydrogen desorption energy from the fluorite MgH2 layer in the multilayers is significantly lower than that of the bulk rutile MgH2 . The structural deformation of the MgH2 layer due to the strain induced by TMH2 is found to be responsible for the destabilization of the Mg-H bond: the more structural deformation, the more destabilization of the Mg-H. Our results provide an important insight for the development of new hydrogen-storage materials with desirable thermodynamic properties.

Thermal conductivity of fresh and irradiated U-Mo fuels

NASA Astrophysics Data System (ADS)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

Thermal conductivity of fresh and irradiated U-Mo fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, the thermal conductivity of fresh dispersion fuel at a temperature of 150°C decreases from 59 W/m ·K down to 18  W/m ·K at a burn-up of 4.9 ·10 21 f/cc and further down to 9 W/m·K at a burn-up of 6.1·10 21 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep as for the dispersion fuel. For a burn-up ofmore » 3.5·10 21 f /cc of monolithic fuel 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. The difference of the decrease of both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increasing burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice affects both dispersion and monolithic fuel.« less

High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

NASA Astrophysics Data System (ADS)

Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

2015-12-01

The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

Effects of thermal treatment on the co-rolled U-Mo fuel foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Tammy L. Trowbridge; Cynthia R. Breckenridge

2014-11-01

A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface betweenmore » the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.« less

Structure of the intermediate Zr/sub 2/Br/sub 2/H by neutron diffraction and its structural and bonding relationships to other phases

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wijeyesekera, S.D.; Corbett, J.D.

1986-12-17

The structures of the isomorphous Zr/sub 2/Br/sub 2/D and Zr/sub 2/Br/sub 2/H have been solved and refined by using Rietveld techniques on pulsed neutron diffraction data obtained from the powdered samples at 14 K (C2/m, a = 19.437 (3) A, b = 3.5253 (4) A, c = 5.9036 (6) A, ..beta.. = 100.98 (1)/sup 0/, R(profile)/R(expected) = 2.44 for the deuteride). The structure consists of layers sequenced Br-Zr-H-H-Zr-Br and arranged such that hydride lies in zigzag chains of distorted metal tetrahedra (or butterflies) (d(Zr-D) = 2.03-2.20 A; d(D-D) = 2.93 A). The structure is intermediate between ZrBr (ccp) and ZrBrHmore » (hcp heavy atoms, double H in trigonal-antiprismatic interstices) and can be generated by concerted intraslab slippage from either. The hemihydride effectively retains most of the strong Zr-Zr bonding of the ZrBr parent while tetrahedral bonding of hydrogen to metal is gained that is absent in ZrBrH. The energetics associated with the contrasting structures of YClH/sub x/ (ZrBr type) and ZrBrH are considered in terms of the results of extended-Hueckel band calculations. 25 references, 7 figures, 3 tables.« less

Metal hydride composition and method of making

DOEpatents

Congdon, James W.

1995-01-01

A dimensionally stable hydride composition and a method for making such a composition. The composition is made by forming particles of a metal hydride into porous granules, mixing the granules with a matrix material, forming the mixture into pellets, and sintering the pellets in the absence of oxygen. The ratio of matrix material to hydride is preferably between approximately 2:1 and 4:1 by volume. The porous structure of the granules accommodates the expansion that occurs when the metal hydride particles absorb hydrogen. The porous matrix allows the flow of hydrogen therethrough to contact the hydride particles, yet supports the granules and contains the hydride fines that result from repeated absorption/desorption cycles.

Cyclopentadiene-mediated hydride transfer from rhodium complexes.

PubMed

Pitman, C L; Finster, O N L; Miller, A J M

2016-07-12

Attempts to generate a proposed rhodium hydride catalytic intermediate instead resulted in isolation of (Cp*H)Rh(bpy)Cl (1), a pentamethylcyclopentadiene complex, formed by C-H bond-forming reductive elimination from the fleeting rhodium hydride. The hydride transfer ability of diene 1 was explored through thermochemistry and hydride transfer reactions, including the reduction of NAD(+).

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sindelar, R.; Louthan, M.; PNNL, B.

2015-05-29

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history,more » residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Method of producing a chemical hydride

DOEpatents

Klingler, Kerry M.; Zollinger, William T.; Wilding, Bruce M.; Bingham, Dennis N.; Wendt, Kraig M.

2007-11-13

A method of producing a chemical hydride is described and which includes selecting a composition having chemical bonds and which is capable of forming a chemical hydride; providing a source of a hydrocarbon; and reacting the composition with the source of the hydrocarbon to generate a chemical hydride.

Fabrication and electrochemical performance of a stable, anode supported thin BaCe0.4Zr0.4Y0.2O3-δ electrolyte Protonic Ceramic Fuel Cell

NASA Astrophysics Data System (ADS)

Nasani, Narendar; Ramasamy, Devaraj; Mikhalev, Sergey; Kovalevsky, Andrei V.; Fagg, Duncan P.

2015-03-01

The present work deals with the fabrication and electrochemical characterisation of a potential protonic ceramic fuel cell based on a Ni-BaZr0.85Y0.15O3-δ anode supported thin film proton conducting BaCe0.4Zr0.4Y0.2O3-δ electrolyte with a Pr2NiO4+δ cathode. Anode and electrolyte materials were prepared by an acetate-H2O2 combustion method. A thin (∼5 μm), dense and crack free BaCe0.4Zr0.4Y0.2O3-δ electrolyte film was successfully obtained on a porous anode support by spin coating and firing at 1450 °C. Maximum power densities of 234, 158, 102 and 63 mW cm-2 at 700, 650, 600 and 550 °C, respectively were achieved for the Ni-BaZr0.85Y0.15O3-δ/BaCe0.4Zr0.4Y0.2O3-δ/Pr2NiO4+δ single cell under fuel cell testing conditions. Electrode polarisation resistance was assessed at open circuit conditions by use of electrochemical impedance spectroscopy (EIS) and is shown to dominate the area specific resistance at low temperatures. Postmortem analysis by scanning electron microscopy (SEM), reveals that no delamination occurs at anode/electrolyte or electrolyte/cathode interfaces upon cell operation.

Metallic glassy Zr70Ni20Pd10 powders for improving the hydrogenation/dehydrogenation behavior of MgH2

PubMed Central

El-Eskandarany, M. Sherif

2016-01-01

Because of its low density, storage of hydrogen in the gaseous and liquids states possess technical and economic challenges. One practical solution for utilizing hydrogen in vehicles with proton-exchange fuel cells membranes is storing hydrogen in metal hydrides. Magnesium hydride (MgH2) remains the best hydrogen storage material due to its high hydrogen capacity and low cost of production. Due to its high activation energy and poor hydrogen sorption/desorption kinetics at moderate temperatures, the pure form of MgH2 is usually mechanically treated by high-energy ball mills and catalyzed with different types of catalysts. These steps are necessary for destabilizing MgH2 to enhance its kinetics behaviors. In the present work, we used a small mole fractions (5 wt.%) of metallic glassy of Zr70Ni20Pd10 powders as a new enhancement agent to improve its hydrogenation/dehydrogenation behaviors of MgH2. This short-range ordered material led to lower the decomposition temperature of MgH2 and its activation energy by about 121 °C and 51 kJ/mol, respectively. Complete hydrogenation/dehydrogenation processes were successfully achieved to charge/discharge about 6 wt.%H2 at 100 °C/200 °C within 1.18 min/3.8 min, respectively. In addition, this new nanocomposite system shows high performance of achieving continuous 100 hydrogen charging/discharging cycles without degradation. PMID:27220994

Hydride heat pump

DOEpatents

Cottingham, James G.

1977-01-01

Method and apparatus for the use of hydrides to exhaust heat from one temperature source and deliver the thermal energy extracted for use at a higher temperature, thereby acting as a heat pump. For this purpose there are employed a pair of hydridable metal compounds having different characteristics working together in a closed pressure system employing a high temperature source to upgrade the heat supplied from a low temperature source.

Metal hydride composition and method of making

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, J.W.

1995-08-22

A dimensionally stable hydride composition and a method for making such a composition are disclosed. The composition is made by forming particles of a metal hydride into porous granules, mixing the granules with a matrix material, forming the mixture into pellets, and sintering the pellets in the absence of oxygen. The ratio of matrix material to hydride is preferably between approximately 2:1 and 4:1 by volume. The porous structure of the granules accommodates the expansion that occurs when the metal hydride particles absorb hydrogen. The porous matrix allows the flow of hydrogen there through to contact the hydride particles, yetmore » supports the granules and contains the hydride fines that result from repeated absorption/desorption cycles. 3 figs.« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE PAGES

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.; ...

2017-03-23

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

Liquid suspensions of reversible metal hydrides

DOEpatents

Reilly, J.J.; Grohse, E.W.; Winsche, W.E.

1983-12-08

The reversibility of the process M + x/2 H/sub 2/ ..-->.. MH/sub x/, where M is a metal hydride former that forms a hydride MH/sub x/ in the presence of H/sub 2/, generally used to store and recall H/sub 2/, is found to proceed under a liquid, thereby to reduce contamination, provide better temperature control and provide in situ mobility of the reactants. Thus, a slurry of particles of a metal hydride former with an inert solvent is subjected to temperature and pressure controlled atmosphere containing H/sub 2/, to store hydrogen (at high pressures) and to release (at low pressures) previously stored hydrogen. The direction of the flow of the H/sub 2/ through the liquid is dependent upon the H/sub 2/ pressure in the gas phase at a given temperature. When the former is above the equilibrium absorption pressure of the respective hydride the reaction proceeds to the right, i.e., the metal hydride is formed and hydrogen is stored in the solid particle. When the H/sub 2/ pressure in the gas phase is below the equilibrium dissociation pressure of the respective hydride the reaction proceeds to the left, the metal hydride is decomposed and hydrogen is released into the gas phase.

Sealed aerospace metal-hydride batteries

NASA Technical Reports Server (NTRS)

Coates, Dwaine

1992-01-01

Nickel metal hydride and silver metal hydride batteries are being developed for aerospace applications. There is a growing market for smaller, lower cost satellites which require higher energy density power sources than aerospace nickel-cadmium at a lower cost than space nickel-hydrogen. These include small LEO satellites, tactical military satellites and satellite constellation programs such as Iridium and Brilliant Pebbles. Small satellites typically do not have the spacecraft volume or the budget required for nickel-hydrogen batteries. NiCd's do not have adequate energy density as well as other problems such as overcharge capability and memory effort. Metal hydride batteries provide the ideal solution for these applications. Metal hydride batteries offer a number of advantages over other aerospace battery systems.

Exsolution lamellae as fast diffusion pathways in rutile: implications for U-Pb thermochronology and Zr thermometry

NASA Astrophysics Data System (ADS)

Smye, A.; Seman, S.; Roberts, N. M. W.; Condon, D. J.; Davis, B.

2017-12-01

Geophysical processes impart characteristic thermal signatures to the lithosphere. Near-continuous thermal histories can be obtained from inversion of intracrystalline U-Pb age profiles in rutile and apatite provided that it can be shown that profile formed in response to Fickian-type diffusion. Here, we present the results of a combined LA-ICPMS and ID-TIMS U-Pb study on rutile grains from two garnet-bearing granulite xenoliths from a kimberlite in the Archean Slave province. Interpreted using numerical models, we show that the rutile U-Pb isotope systematics are consistent with slow-cooling following crystallization at 1.2 Ga, contemporaneous with the Mackenzie dike swarm. However, inversion of rutile U-Pb age gradients is complicated by the ubiquitous presence of ilmenite exsolution lamellae. We show that these lamellae act as fast diffusion pathways for Pb and High Field Strength Elements, including Zr. This has important implications for the use of rutile as a U-Pb themochronometer and as a single-phase thermometer.

Direct synthesis of catalyzed hydride compounds

DOEpatents

Gross, Karl J.; Majzoub, Eric

2004-09-21

A method is disclosed for directly preparing alkali metal aluminum hydrides such as NaAlH.sub.4 and Na.sub.3 AlH.sub.6 from either the alkali metal or its hydride, and aluminum. The hydride thus prepared is doped with a small portion of a transition metal catalyst compound, such as TiCl.sub.3, TiF.sub.3, or a mixture of these materials, in order to render them reversibly hydridable. The process provides for mechanically mixing the dry reagents under an inert atmosphere followed by charging the mixed materials with high pressure hydrogen while heating the mixture to about 125.degree. C. The method is relatively simple and inexpensive and provides reversible hydride compounds which are free of the usual contamination introduced by prior art wet chemical methods.

Zr-92(d,p)Zr-93 and Zr-92(d,t)Zr-91

NASA Technical Reports Server (NTRS)

Baron, N.; Fink, C. L.; Christensen, P. R.; Nickels, J.; Torsteinsen, T.

1972-01-01

The structures of Zr-93 and Zr-91 were studied by the stripping reaction Zr-92(d,p)Zr-93 and the pick-up reaction Zr-92(d,t)Zr-91 using 13 MeV incident deuterons. The reaction product particles were detected by counter telescope. Typical spectra from the reactions were analyzed by a nonlinear least squares peak fitting program which included a background search. Spin and parity assignments to observed excited levels were made by comparing experimental angular distributions with distorted wave Born approximation calculations.

Hydrogenation behavior of Ti-implanted Zr-1Nb alloy with TiN films deposited using filtered vacuum arc and magnetron sputtering

NASA Astrophysics Data System (ADS)

Kashkarov, E. B.; Nikitenkov, N. N.; Sutygina, A. N.; Bezmaternykh, A. O.; Kudiiarov, V. N.; Syrtanov, M. S.; Pryamushko, T. S.

2018-02-01

More than 60 years of operation of water-cooled reactors have shown that local or general critical hydrogen concentration is one of the basic limiting criteria of zirconium-based fuel element claddings. During the coolant radiolysis, released hydrogen penetrates and accumulates in zirconium alloys. Hydrogenation of zirconium alloys leads to degradation of their mechanical properties, hydride cracking and stress corrosion cracking. In this research the effect of titanium nitride (TiN) deposition on hydrogenation behavior of Ti-implanted Zr-1Nb alloy was described. Ti-implanted interlayer was fabricated by plasma immersion ion implantation (PIII) at the pulsed bias voltage of 1500 V to improve the adhesion of TiN and reduce hydrogen penetration into Zr-1Nb alloy. We conducted the comparative analysis on hydrogenation behavior of the Ti-implanted alloy with sputtered and evaporated TiN films by reactive dc magnetron sputtering (dcMS) and filtered cathodic vacuum arc deposition (FVAD), respectively. The crystalline structure and surface morphology were investigated using X-ray diffraction (XRD) and scanning electron microscopy (SEM). The elemental distribution was analyzed using glow-discharge optical emission spectroscopy (GD-OES). Hydrogenation was performed from gas atmosphere at 350 °C and 2 atm hydrogen pressure. The results revealed that TiN films as well as Ti implantation significantly reduce hydrogen absorption rate of Zr-1Nb alloy. The best performance to reduce the rate of hydrogen absorption is Ti-implanted layer with evaporated TiN film. Morphology of the films impacted hydrogen permeation through TiN films: the denser film the lower hydrogen permeation. The Ti-implanted interface plays an important role of hydrogen accumulation layer for trapping the penetrated hydrogen. No deterioration of adhesive properties of TiN films on Zr-1Nb alloy with Ti-implanted interface occurs under high-temperature hydrogen exposure. Thus, the fabrication of Ti

Irradiation behavior of U 6Mn-Al dispersion fuel elements

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Modeling a failure criterion for U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore » fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to

Hydrogen Storage Engineering Center of Excellence Metal Hydride Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motyka, T.

2014-05-31

The Hydrogen Storage Engineering Center of Excellence (HSECoE) was established in 2009 by the U.S. Department of Energy (DOE) to advance the development of materials-based hydrogen storage systems for hydrogen-fueled light-duty vehicles. The overall objective of the HSECoE is to develop complete, integrated system concepts that utilize reversible metal hydrides, adsorbents, and chemical hydrogen storage materials through the use of advanced engineering concepts and designs that can simultaneously meet or exceed all the DOE targets. This report describes the activities and accomplishments during Phase 1 of the reversible metal hydride portion of the HSECoE, which lasted 30 months from Februarymore » 2009 to August 2011. A complete list of all the HSECoE partners can be found later in this report but for the reversible metal hydride portion of the HSECoE work the major contributing organizations to this effort were the United Technology Research Center (UTRC), General Motors (GM), Pacific Northwest National Laboratory (PNNL), the National Renewable Energy Laboratory (NREL) and the Savannah River National Laboratory (SRNL). Specific individuals from these and other institutions that supported this effort and the writing of this report are included in the list of contributors and in the acknowledgement sections of this report. The efforts of the HSECoE are organized into three phases each approximately 2 years in duration. In Phase I, comprehensive system engineering analyses and assessments were made of the three classes of storage media that included development of system level transport and thermal models of alternative conceptual storage configurations to permit detailed comparisons against the DOE performance targets for light-duty vehicles. Phase 1 tasks also included identification and technical justifications for candidate storage media and configurations that should be capable of reaching or exceeding the DOE targets. Phase 2 involved bench-level testing and

Multiphysics phase field modeling of hydrogen diffusion and delta-hydride precipitation in alpha-zirconium

NASA Astrophysics Data System (ADS)

Jokisaari, Andrea M.

Hydride precipitation in zirconium is a significant factor limiting the lifetime of nuclear fuel cladding, because hydride microstructures play a key role in the degradation of fuel cladding. However, the behavior of hydrogen in zirconium has typically been modeled using mean field approaches, which do not consider microstructural evolution. This thesis describes a quantitative microstructural evolution model for the alpha-zirconium/delta-hydride system and the associated numerical methods and algorithms that were developed. The multiphysics, phase field-based model incorporates CALPHAD free energy descriptions, linear elastic solid mechanics, and classical nucleation theory. A flexible simulation software implementing the model, Hyrax, is built on the Multiphysics Object Oriented Simulation Environment (MOOSE) finite element framework. Hyrax is open-source and freely available; moreover, the numerical methods and algorithms that have been developed are generalizable to other systems. The algorithms are described in detail, and verification studies for each are discussed. In addition, analyses of the sensitivity of the simulation results to the choice of numerical parameters are presented. For example, threshold values for the CALPHAD free energy algorithm and the use of mesh and time adaptivity when employing the nucleation algorithm are studied. Furthermore, preliminary insights into the nucleation behavior of delta-hydrides are described. These include a) the sensitivities of the nucleation rate to temperature, interfacial energy, composition and elastic energy, b) the spatial variation of the nucleation rate around a single precipitate, and c) the effect of interfacial energy and nucleation rate on the precipitate microstructure. Finally, several avenues for future work are discussed. Topics encompass the terminal solid solubility hysteresis of hydrogen in zirconium and the effects of the alpha/delta interfacial energy, as well as thermodiffusion, plasticity

Calculation of thermodynamic hydricities and the design of hydride donors for CO2 reduction

PubMed Central

Muckerman, James T.; Achord, Patrick; Creutz, Carol; Polyansky, Dmitry E.; Fujita, Etsuko

2012-01-01

We have developed a correlation between experimental and density functional theory-derived results of the hydride-donating power, or “hydricity”, of various ruthenium, rhenium, and organic hydride donors. This approach utilizes the correlation between experimental hydricity values and their corresponding calculated free-energy differences between the hydride donors and their conjugate acceptors in acetonitrile, and leads to an extrapolated value of the absolute free energy of the hydride ion without the necessity to calculate it directly. We then use this correlation to predict, from density functional theory-calculated data, hydricity values of ruthenium and rhenium complexes that incorporate the pbnHH ligand—pbnHH = 1,5-dihydro-2-(2-pyridyl)-benzo[b]-1,5-naphthyridine—to model the function of NADPH. These visible light-generated, photocatalytic complexes produced by disproportionation of a protonated-photoreduced dimer of a metal-pbn complex may be valuable for use in reducing CO2 to fuels such as methanol. The excited-state lifetime of photoexcited [Ru(bpy)2(pbnHH)]2+ is found to be about 70 ns, and this excited state can be reductively quenched by triethylamine or 1,4-diazabicyclo[2.2.2]octane to produce the one-electron-reduced [Ru(bpy)2(pbnHH)]+ species with half-life exceeding 50 μs, thus opening the door to new opportunities for hydride-transfer reactions leading to CO2 reduction by producing a species with much increased hydricity. PMID:22826261

Metal hydride compositions and lithium ion batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Young, Kwo; Nei, Jean

Heterogeneous metal hydride (MH) compositions comprising a main region comprising a first metal hydride and a secondary region comprising one or more additional components selected from the group consisting of second metal hydrides, metals, metal alloys and further metal compounds are suitable as anode materials for lithium ion cells. The first metal hydride is for example MgH.sub.2. Methods for preparing the composition include coating, mechanical grinding, sintering, heat treatment and quenching techniques.

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Keiser, D. D.; Miller, B. D.

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE PAGES

Gan, J.; Keiser, D. D.; Miller, B. D.; ...

2017-07-15

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Kazimi, Mujid S.

2013-10-01

The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Examination of Multiphase (Zr,Ti)(V,Cr,Mn,Ni)2 Ni-MH Electrode Alloys: Part I. Dendritic Solidification Structure

NASA Astrophysics Data System (ADS)

Boettinger, W. J.; Newbury, D. E.; Wang, K.; Bendersky, L. A.; Chiu, C.; Kattner, U. R.; Young, K.; Chao, B.

2010-08-01

The solidification microstructures of three nine-element Zr-Ni-based AB2 type C14/C15 Laves hydrogen storage alloys are determined. The selected compositions represent a class of alloys being examined for usage as an MH electrode in nickel metal-hydride batteries that often have their best properties in the cast state. Solidification is accomplished by dendritic growth of hexagonal C14 Laves phase, peritectic solidification of cubic C15 Laves phase, and formation of cubic B2 phase in the interdendritic regions. The B2 phase decomposes in the solid state into a complex multivariate platelike structure containing Zr-Ni-rich intermetallics. The observed sequence C14/C15 upon solidification agrees with predictions using effective compositions and thermodynamic assessments of the ternary systems, Ni-Cr-Zr and Cr-Ti-Zr. Experimentally, the closeness of the compositions of the C14 and C15 phases required the use of compositional mapping with an energy dispersive detector capable of processing a very high X-ray flux to locate regions in the microstructure for quantitative composition measurement and transmission electron microscope examination.

Metal Hydrides for High-Temperature Power Generation

DOE PAGES

Ronnebro, Ewa; Whyatt, Greg A.; Powell, Michael R.; ...

2015-08-10

Metal hydrides can be utilized for hydrogen storage and for thermal energy storage (TES) applications. By using TES with solar technologies, heat can be stored from sun energy to be used later which enables continuous power generation. We are developing a TES technology based on a dual-bed metal hydride system, which has a high-temperature (HT) metal hydride operating reversibly at 600-800°C to generate heat as well as a low-temperature (LT) hydride near room temperature that is used for hydrogen storage during sun hours until there is a need to produce electricity, such as during night time, a cloudy day, ormore » during peak hours. We proceeded from selecting a high-energy density, low-cost HT-hydride based on performance characterization on gram size samples, to scale-up to kilogram quantities and design, fabrication and testing of a 1.5kWh, 200kWh/m 3 bench-scale TES prototype based on a HT-bed of titanium hydride and a hydrogen gas storage instead of a LT-hydride. COMSOL Multiphysics was used to make performance predictions for cylindrical hydride beds with varying diameters and thermal conductivities. Based on experimental and modeling results, a bench-scale prototype was designed and fabricated and we successfully showed feasibility to meet or exceed all performance targets.« less

High density dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heuser, Brent; Stubbins, James; Kozlowski, Tomasz

The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys.more » The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and

Examination of Multiphase (Zr,Ti)(V,Cr,Mn,Ni)2 Ni-MH Electrode Alloys: Part II. Solid-State Transformation of the Interdendritic B2 Phase

NASA Astrophysics Data System (ADS)

Bendersky, L. A.; Wang, K.; Boettinger, W. J.; Newbury, D. E.; Young, K.; Chao, B.

2010-08-01

Solidification microstructure of multicomponent (Zr,Ti)-Ni-(V,Cr,Mn,Co) alloys intended for use as negative electrodes in Ni-metal hydride (Ni-MH) batteries was studied in Part I of this series of articles. Part II of the series examines the complex internal structure of the interdendritic grains formed by solid-state transformation and believed to play an important role in the electrochemical charge/discharge characteristics of the overall alloy composition. By studying one alloy, Zr21Ti12.5V10Cr5.5Mn5.1Co5.0Ni40.2Al0.5Sn0.3, it is shown that the interdendritic grains solidify as a B2 (Ti,Zr)44(Ni,TM)56 phase, and then undergo transformation to Zr7Ni10-type, Zr9Ni11-type, and martensitic phases. The transformations obey orientation relationships between the high-temperature B2 phase and the low-temperature Zr-Ni-type intermetallics, and consequently lead to a multivariant structure. The major orientation relationship for the orthorhombic Zr7Ni10 type is [011]Zr7Ni10//[001]B2; (100)Zr7Ni10//(100)B2. The orientation relationship for the tetragonal Zr9Ni11 type is [001]Zr9Ni11//[001]B2; (130)Zr9Ni11//(100)B2. Binary Ni-Zr and ternary Ti-Ni-Zr phase diagrams were used to rationalize the formation of the observed domain structure.

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, andmore » reduced the volume expansion.« less

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Thermal coupling potential of Solid Oxide Fuel Cells with metal hydride tanks: Thermodynamic and design considerations towards integrated systems

NASA Astrophysics Data System (ADS)

Yiotis, Andreas G.; Kainourgiakis, Michael E.; Kosmidis, Lefteris I.; Charalambopoulou, Georgia C.; Stubos, Athanassios K.

2014-12-01

We study the thermal coupling potential between a high temperature metal hydride (MH) tank and a Solid Oxide Fuel Cell (SOFC) aiming towards the design of an efficient integrated system, where the thermal power produced during normal SOFC operation is redirected towards the MH tank in order to maintain H2 desorption without the use of external heating sources. Based on principles of thermodynamics, we calculate the energy balance in the SOFC/MH system and derive analytical expressions for both the thermal power produced during SOFC operation and the corresponding thermal power required for H2 desorption, as a function of the operating temperature, efficiency and fuel utilization ratio in the SOFC, and the MH enthalpy of desorption in the tank. Based on these calculations, we propose an integrated SOFC/MH design where heat is transferred primarily by radiation to the tank in order to maintain steady-state desorption conditions. We develop a mathematical model for this particular design that accounts for heat/mass transfer and desorption kinetics in the tank, and solve for the dynamics of the system assuming MgH2 as a storage material. Our results focus primarily on tank operating conditions, such as pressure, temperature and H2 saturation profiles vs operation time.

Iron hydrides formation in interstellar clouds

NASA Astrophysics Data System (ADS)

Bar-Nun, A.; Pasternak, M.; Barrett, P. H.

1980-07-01

A recent Moessbauer study with Fe-57 in a solid hydrogen or hydrogen-argon matrix demonstrated the formation of an iron hydride molecule (FeH2) at 2.5-5 K. Following this and other studies, the possible existence of iron hydride molecules in interstellar clouds is proposed. In clouds, the iron hydrides FeH and FeH2 would be formed only on grains, by encounters of H atoms or H2 molecules with Fe atoms which are adsorbed on the grains. The other transition metals, Sc, Ti, V, Cr, Mn, Co, N, Cd and also Cu and Ca form hydrides of the type M-H, which could be responsible, at least in part, for the depletion of these metals in clouds.

A study of advanced magnesium-based hydride and development of a metal hydride thermal battery system

NASA Astrophysics Data System (ADS)

Zhou, Chengshang

Metal hydrides are a group of important materials known as energy carriers for renewable energy and thermal energy storage. A concept of thermal battery based on advanced metal hydrides is studied for heating and cooling of cabins in electric vehicles. The system utilizes a pair of thermodynamically matched metal hydrides as energy storage media. The hot hydride that is identified and developed is catalyzed MgH2 due to its high energy density and enhanced kinetics. TiV0.62Mn1.5, TiMn2, and LaNi5 alloys are selected as the matching cold hydride. A systematic experimental survey is carried out in this study to compare a wide range of additives including transitions metals, transition metal oxides, hydrides, intermetallic compounds, and carbon materials, with respect to their effects on dehydrogenation properties of MgH2. The results show that additives such as Ti and V-based metals, hydride, and certain intermetallic compounds have strong catalytic effects. Solid solution alloys of magnesium are exploited as a way to destabilize magnesium hydride thermodynamically. Various elements are alloyed with magnesium to form solid solutions, including indium and aluminum. Thermodynamic properties of the reactions between the magnesium solid solution alloys and hydrogen are investigated, showing that all the solid solution alloys that are investigated in this work have higher equilibrium hydrogen pressures than that of pure magnesium. Cyclic stability of catalyzed MgH2 is characterized and analyzed using a PCT Sievert-type apparatus. Three systems, including MgH2-TiH 2, MgH2-TiMn2, and MgH2-VTiCr, are examined. The hydrogenating and dehydrogenating kinetics at 300°C are stable after 100 cycles. However, the low temperature (25°C to 150°C) hydrogenation kinetics suffer a severe degradation during hydrogen cycling. Further experiments confirm that the low temperature kinetic degradation can be mainly related the extended hydrogenation-dehydrogenation reactions. Proof

Mechanistic insights into iron catalyzed dehydrogenation of formic acid: β-hydride elimination vs. direct hydride transfer.

PubMed

Yang, Xinzheng

2013-09-07

Density functional theory calculations reveal a complete reaction mechanism with detailed energy profiles and transition state structures for the dehydrogenation of formic acid catalyzed by an iron complex, [P(CH2CH2PPh2)3FeH](+). In the cationic reaction pathway, a β-hydride elimination process is confirmed to be the rate-determining step in this catalytic reaction. A potential reaction pathway starting with a direct hydride transfer from HCOO(-) to Fe is found to be possible, but slightly less favorable than the catalytic cycle with a β-hydride elimination step.

Chemical reactivity testing for the National Spent Nuclear Fuel Program. Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koester, L.W.

This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, Y60-101PD, Quality Program Description, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will bemore » noted. The project consists of conducting three separate series of related experiments, ''Passivation of Uranium Hydride Powder With Oxygen and Water'', '''Passivation of Uranium Hydride Powder with Surface Characterization'', and ''Electrochemical Measure of Uranium Hydride Corrosion Rate''.« less

Interdiffusion in U 3Si-Al, U 3Si 2-Al, and USi-Al dispersion fuels during irradiation

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, Gerard L.

2011-03-01

Uranium-silicide compound fuel dispersion in an Al matrix is used in research and test reactors worldwide. Interaction layer (IL) growth between fuel particles and the matrix is one of performance issues. The interaction layer growth data for U 3Si, U 3Si 2 and USi dispersions in Al were obtained from both out-of-pile and in-pile tests. The IL is dominantly U(AlSi) 3 from out-of-pile tests, but its (Al + Si)/U ratio from in-pile tests is higher than the out-of-pile data, because of amorphous behavior of the ILs. IL growth correlations were developed for U 3Si-Al and U 3Si 2-Al. The IL growth rates were dependent on the U/Si ratio of the fuel compounds. During irradiation, however, the IL growth rates did not decrease with the decreasing U/Si ratio by fission. It is reasoned that transition metal fission products in the IL compensate the loss of U atoms by providing chemical potential for Al diffusion and volume expansion by solid swelling and gas bubble swelling. The addition of Mo in U 3Si 2 reduces the IL growth rate, which is similar to that of UMo alloy dispersion in a silicon-added Al matrix.

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Postirradiation analysis of the latest high uranium density miniplate test: RERTR 8.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hofman, G. L.; Kim, Y. S.; Rest, J.

2008-01-01

Results of destructive examination of fuel miniplates irradiated in the RERTR-8 test are discussed. Metallographic features of dispersion fuel containing fuel particles of U-7wt%Mo with 1wt% Ti or 2wt% Zr are analyzed. It is hypothesized that Zr, either as alloy addition or fission product, may have a destabilizing effect on fission gas behavior. The purpose of miniplate test RERTR-8 was to obtain irradiation performance data on monolithic fuel plates fabricated by friction bonding (FB) and isostatic hot pressing (HIP), as well as dispersion fuel plates that contain U-7Mo fuel particles alloyed with small amounts of Zr or Ti (see Fig.more » 1). The results of the monolithic plates destructively examined to date were presented at the 2007 RERTR meeting in Prague. This paper presents the first results on the dispersion plates with Ti and Zr additions to U-7Mo. The effect of Ti and Zr additions to U-7wt%Mo on the extent of fuel-aluminum interdiffusion, although measureable, is small in absolute terms because of the overwhelming effect of the 5% Si addition to the Al matrix. Ti additions to the U-7wt%Mo have no discernable effect on swelling behavior of the fuel. However, there are indications that the addition of Zr may have a destabilizing effect on fission gas behavior at high burnup.« less

Silanes as Fuel for Aerospace Propulsion

NASA Astrophysics Data System (ADS)

Simone, Domenico; Bruno, Claudio; Hidding, Bernhard

In the light of recently revived interest in high energy density fuels for aerospace applications1,2), a new look is being given at unconventional fuels. Among the latter are hydrides, because their hydrogen content and density. Among hydrides silanes are of interest because of their combustion and energetic properties.Silanes are silicon hydrides organized in molecular chains similar to those of hydrocarbons; at STP, lower silanes (SiH4, Si2H6) are gaseous and extremely pyrophoric; with increasing chain length, silanes become liquid from trisilane (Si3H8) on, and therefore easily pumped. Another important feature of silanes is the large amount of hydrogen theoretically available by thermal decomposition: in fact at moderate temperatures (about 500 K) the chains begin to break and at 700 K their decomposition is complete, yielding silicon and gaseous hydrogen, useful for propulsion in combination with air nitrogen and oxygen. This last feature, if confirmed, could identify silanes not only as energy carriers but also components in bi-fuel systems. To assess their theoretical performance, simulations were conducted assuming silanes and/or their thermal decomposition products in combination with various oxidizers and air. Preliminary results are suggestive of their potential for some specialized applications, especially where compactness is at premium.

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

2016-12-01

The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

Verification and Validation Strategy for Implementation of Hybrid Potts-Phase Field Hydride Modeling Capability in MBM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason D. Hales; Veena Tikare

2014-04-01

The Used Fuel Disposition (UFD) program has initiated a project to develop a hydride formation modeling tool using a hybrid Potts­phase field approach. The Potts model is incorporated in the SPPARKS code from Sandia National Laboratories. The phase field model is provided through MARMOT from Idaho National Laboratory.

FY 2016 Status Report: Documentation of All CIRFT Data including Hydride Reorientation Tests (Draft M2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

The first portion of this report provides a detailed description of fiscal year (FY) 2015 test result corrections and analysis updates based on FY 2016 updates to the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) program methodology, which is used to evaluate the vibration integrity of spent nuclear fuel (SNF) under normal conditions of transport (NCT). The CIRFT consists of a U-frame test setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages connecting to a universal testing machine. The curvature SNF rod bending is obtained through a three-point deflection measurementmore » method. Three linear variable differential transformers (LVDTs) are clamped to the side connecting plates of the U-frame and used to capture deformation of the rod. The second portion of this report provides the latest CIRFT data, including data for the hydride reorientation test. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNF samples into a single zone. A detailed examination revealed that, at the same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), LMK, and mixed uranium-plutonium oxide (MOX). Just looking at the strain, LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The third portion of this report provides finite element analysis (FEA) dynamic deformation simulation of SNF assemblies . In a horizontal layout under NCT, the fuel assembly’s skeleton, which is formed by guide tubes and spacer grids, is the primary load bearing apparatus carrying and transferring vibration loads within an SNF assembly. These vibration loads include interaction forces between the SNF assembly and the canister basket walls. Therefore, the integrity of the

Irradiation performance of HTGR fuel rods in HFIR experiments HRB-7 and -8

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valentine, K.H.; Homan, F.J.; Long, E.L. Jr.

1977-05-01

The HRB-7 and -8 experiments were designed as a comprehensive test of mixed thorium-uranium oxide fissile particles with Th:U ratios from 0 to 8 for HTGR recycle application. In addition, fissile particles derived from Weak-Acid Resin (WAR) were tested as a potential backup type of fissile particle for HTGR recycle. These experiments were conducted at two temperatures (1250 and 1500/sup 0/C) to determine the influence of operating temperature on the performance parameters studied. The minor objectives were comparison of advanced coating designs where ZrC replaced SiC in the Triso design, testing of fuel coated in laboratory-scale equipment with fuel coatedmore » in production-scale coaters, comparison of the performance of /sup 233/U-bearing particles with that of /sup 235/U-bearing particles, comparison of the performance of Biso coatings with Triso coatings for particles containing the same type of kernel, and testing of multijunction tungsten-rhenium thermocouples. All objectives were accomplished. As a result of these experiments the mixed thorium-uranium oxide fissile kernel was replaced by a WAR-derived particle in the reference recycle design. A tentative decision to make this change had been reached before the HRB-7 and -8 capsules were examined, and the results of the examination confirmed the accuracy of the previous decision. Even maximum dilution (Th/U approximately equal to 8) of the mixed thorium-uranium oxide kernel was insufficient to prevent amoeba of the kernels at rates that are unacceptable in a large HTGR. Other results showed the performance of /sup 233/U-bearing particles to be identical to that of /sup 235/U-bearing particles, the performance of fuel coated in production-scale equipment to be at least as good as that of fuel coated in laboratory-scale coaters, the performance of ZrC coatings to be very promising, and Biso coatings to be inferior to Triso coatings relative to fission product retention.« less

Mechanical and thermal properties of bulk ZrB2

NASA Astrophysics Data System (ADS)

Nakamori, Fumihiro; Ohishi, Yuji; Muta, Hiroaki; Kurosaki, Ken; Fukumoto, Ken-ichi; Yamanaka, Shinsuke

2015-12-01

ZrB2 appears to have formed in the fuel debris at the Fukushima Daiichi nuclear disaster site, through the reaction between Zircaloy cladding materials and the control rod material B4C. Since ZrB2 has a high melting point of 3518 K, the ceramic has been widely studied as a heat-resistant material. Although various studies on the thermochemical and thermophysical properties have been performed for ZrB2, significant differences exist in the data, possibly due to impurities or the porosity within the studied samples. In the present study, we have prepared a ZrB2 bulk sample with 93.1% theoretical density by sintering ZrB2 powder. On this sample, we have comprehensively examined the thermal and mechanical properties of ZrB2 by the measurement of specific heat, ultrasonic sound velocities, thermal diffusivity, and thermal expansion. Vickers hardness and fracture toughness were also measured and found to be 13-23 GPa and 1.8-2.8 MPa m0.5, respectively. The relationships between these properties were carefully examined in the present study.

A classical but new kinetic equation for hydride transfer reactions.

PubMed

Zhu, Xiao-Qing; Deng, Fei-Huang; Yang, Jin-Dong; Li, Xiu-Tao; Chen, Qiang; Lei, Nan-Ping; Meng, Fan-Kun; Zhao, Xiao-Peng; Han, Su-Hui; Hao, Er-Jun; Mu, Yuan-Yuan

2013-09-28

A classical but new kinetic equation to estimate activation energies of various hydride transfer reactions was developed according to transition state theory using the Morse-type free energy curves of hydride donors to release a hydride anion and hydride acceptors to capture a hydride anion and by which the activation energies of 187 typical hydride self-exchange reactions and more than thirty thousand hydride cross transfer reactions in acetonitrile were safely estimated in this work. Since the development of the kinetic equation is only on the basis of the related chemical bond changes of the hydride transfer reactants, the kinetic equation should be also suitable for proton transfer reactions, hydrogen atom transfer reactions and all the other chemical reactions involved with breaking and formation of chemical bonds. One of the most important contributions of this work is to have achieved the perfect unity of the kinetic equation and thermodynamic equation for hydride transfer reactions.

Inhibited solid propellant composition containing beryllium hydride

NASA Technical Reports Server (NTRS)

Thompson, W. W. (Inventor)

1978-01-01

An object of this invention is to provide a composition of beryllium hydride and carboxy-terminated polybutadiene which is stable. Another object of this invention is to provide a method for inhibiting the reactivity of beryllium hydride toward carboxy-terminated polybutadiene. It was found that a small amount of lecithin inhibits the reaction of beryllium hydride with the acid groups in carboxy terminated polybutadiene.

High- and low-Am RE inclusion phases in a U-Np-Pu-Am-Zr alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn E.; Madden, James W.; O'Holleran, Thomas P.

2015-03-01

Structural, microstructural, and microchemical data were collected from rare-earth inclusions in an as-cast U-Pu-Zr alloy with ~3 at% Am, 2% Np, and 9% rare-earth elements (La, Ce, Pr, and Nd). Two RE phases with different concentrations of Am were identified. The composition of high-Am RE inclusions is ~2-5 at% La, 15-20 % Ce, 5-10% Pr, 25-45% Nd, 1% Np, 5-10% Pu, and 10-20% Am. Some areas also have O, although this does not appear to be an essential part of the high-Am RE phase. The inclusions have a face-centered cubic structure with a lattice parameter a ~ 0.54 nm. Themore » composition of the only low-Am RE inclusion studied in detail is ~~35-40 at% O, 40-45 % Nd, 1-2% Zr, 4-5% La, 9-10% Ce, and 6-7% Pr. This inclusion is an oxide with a crystal structure similar to the room-temperature structure of Nd 2O 3. Microstructural features suggest that oxidation occurred during casting, and that early crystallization of high-temperature oxides led to formation of two distinct RE phases.« less

Reactive sputtering of δ-ZrH{sub 2} thin films by high power impulse magnetron sputtering and direct current magnetron sputtering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Högberg, Hans, E-mail: hans.hogberg@liu.se; Tengdelius, Lina; Eriksson, Fredrik

2014-07-01

Reactive sputtering by high power impulse magnetron sputtering (HiPIMS) and direct current magnetron sputtering (DCMS) of a Zr target in Ar/H{sub 2} plasmas was employed to deposit Zr-H films on Si(100) substrates, and with H content up to 61 at. % and O contents typically below 0.2 at. % as determined by elastic recoil detection analysis. X-ray photoelectron spectroscopy reveals a chemical shift of ∼0.7 eV to higher binding energies for the Zr-H films compared to pure Zr films, consistent with a charge transfer from Zr to H in a zirconium hydride. X-ray diffraction shows that the films are single-phase δ-ZrH{sub 2} (CaF{submore » 2} type structure) at H content >∼55 at. % and pole figure measurements give a 111 preferred orientation for these films. Scanning electron microscopy cross-section images show a glasslike microstructure for the HiPIMS films, while the DCMS films are columnar. Nanoindentation yield hardness values of 5.5–7 GPa for the δ-ZrH{sub 2} films that is slightly harder than the ∼5 GPa determined for Zr films and with coefficients of friction in the range of 0.12–0.18 to compare with the range of 0.4–0.6 obtained for Zr films. Wear resistance testing show that phase-pure δ-ZrH{sub 2} films deposited by HiPIMS exhibit up to 50 times lower wear rate compared to those containing a secondary Zr phase. Four-point probe measurements give resistivity values in the range of ∼100–120 μΩ cm for the δ-ZrH{sub 2} films, which is slightly higher compared to Zr films with values in the range 70–80 μΩ cm.« less

49 CFR 173.311 - Metal hydride storage systems.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Metal hydride storage systems. 173.311 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Gases; Preparation and Packaging § 173.311 Metal hydride storage systems. The following packing instruction is applicable to transportable UN Metal hydride storage systems...

49 CFR 173.311 - Metal hydride storage systems.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 2 2011-10-01 2011-10-01 false Metal hydride storage systems. 173.311 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Gases; Preparation and Packaging § 173.311 Metal hydride storage systems. The following packing instruction is applicable to transportable UN Metal hydride storage systems...

49 CFR 173.311 - Metal hydride storage systems.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 2 2014-10-01 2014-10-01 false Metal hydride storage systems. 173.311 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Gases; Preparation and Packaging § 173.311 Metal hydride storage systems. The following packing instruction is applicable to transportable UN Metal hydride storage systems...

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

STAN, MARIUS; HECKER, SIEGFRIED S.

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuelsmore » suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.« less

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily duemore » to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.« less

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

2016-05-01

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Activated aluminum hydride hydrogen storage compositions and uses thereof

DOEpatents

Sandrock, Gary; Reilly, James; Graetz, Jason; Wegrzyn, James E.

2010-11-23

In one aspect, the invention relates to activated aluminum hydride hydrogen storage compositions containing aluminum hydride in the presence of, or absence of, hydrogen desorption stimulants. The invention particularly relates to such compositions having one or more hydrogen desorption stimulants selected from metal hydrides and metal aluminum hydrides. In another aspect, the invention relates to methods for generating hydrogen from such hydrogen storage compositions.

METHOD AND APPARATUS FOR MAKING URANIUM-HYDRIDE COMPACTS

DOEpatents

Wellborn, W.; Armstrong, J.R.

1959-03-10

A method and apparatus are presented for making compacts of pyrophoric hydrides in a continuous operation out of contact with air. It is particularly useful for the preparation of a canned compact of uranium hydride possessing high density and purity. The metallic uranium is enclosed in a container, positioned in a die body evacuated and nvert the uranium to the hydride is admitted and the container sealed. Heat is applied to bring about the formation of the hydride, following which compression is used to form the compact sealed in a container ready for use.

Electrochemical research in chemical hydrogen storage materials: Sodium borohydride and organotin hydrides

NASA Astrophysics Data System (ADS)

McLafferty, Jason

Chemical storage of hydrogen involves release of hydrogen in a controlled manner from materials in which the hydrogen is covalently bound. Sodium borohydride and aminoborane are two materials given consideration as chemical hydrogen storage materials by the US Department of Energy. A very significant barrier to adoption of these materials as hydrogen carriers is their regeneration from "spent fuel," i.e., the material remaining after discharge of hydrogen. In this thesis, some research directed at regeneration of sodium borohydride and aminoborane is described. For sodium borohydride, electrochemical reduction of boric acid and sodium metaborate (representing spent fuel) in alkaline, aqueous solution has been investigated. Similarly to literature reports (primarily patents), a variety of cathode materials were tried in these experiments. Additionally, approaches directed at overcoming electrostatic repulsion of borate anion from the cathode, not described in the previous literature for electrochemical reduction of spent fuels, have been attempted. A quantitative analytical method for measuring the concentration of sodium borohydride in alkaline aqueous solution has been developed as part of this work and is described herein. Finally, findings from stability tests for sodium borohydride in aqueous solutions of several different compositions are reported. For aminoborane, other research institutes have developed regeneration schemes involving tributyltin hydride. In this thesis, electrochemical reduction experiments attempting to regenerate tributyltin hydride from tributyltin chloride (a representative by-product of the regeneration scheme) are described. These experiments were performed in the non-aqueous solvents acetonitrile and 1,2-dimethoxyethane. A non-aqueous reference electrode for electrolysis experiments in acetonitrile was developed and is described.

Fuel cells for low power applications

NASA Astrophysics Data System (ADS)

Heinzel, A.; Hebling, C.; Müller, M.; Zedda, M.; Müller, C.

Electronic devices show an ever-increasing power demand and thus, require innovative concepts for power supply. For a wide range of power and energy capacity, membrane fuel cells are an attractive alternative to conventional batteries. The main advantages are the flexibility with respect to power and capacity achievable with different devices for energy conversion and energy storage, the long lifetime and long service life, the good ecological balance, very low self-discharge. Therefore, the development of fuel cell systems for portable electronic devices is an attractive, although also a challenging, goal. The fuel for a membrane fuel cell might be hydrogen from a hydride storage system or methanol/water as a liquid alternative. The main differences between the two systems are the much higher power density for hydrogen fuel cells, the higher energy density per weight for the liquid fuel, safety aspects and infrastructure for fuel supply for hydride materials. For different applications, different system designs are required. High power cells are required for portable computers, low power methanol fuel cells required for mobile phones in hybrid systems with batteries and micro-fuel cells are required, e.g. for hand held PCs in the sub-Watt range. All these technologies are currently under development. Performance data and results of simulations and experimental investigations will be presented.

Use of reversible hydrides for hydrogen storage

NASA Technical Reports Server (NTRS)

Darriet, B.; Pezat, M.; Hagenmuller, P.

1980-01-01

The addition of metals or alloys whose hydrides have a high dissociation pressure allows a considerable increase in the hydrogenation rate of magnesium. The influence of temperature and hydrogen pressure on the reaction rate were studied. Results concerning the hydriding of magnesium rich alloys such as Mg2Ca, La2Mg17 and CeMg12 are presented. The hydriding mechanism of La2Mg17 and CeMg12 alloys is given.

Hydride affinity scale of various substituted arylcarbeniums in acetonitrile.

PubMed

Zhu, Xiao-Qing; Wang, Chun-Hua

2010-12-23

Combined with the integral equation formalism polarized continuum model (IEFPCM), the hydride affinities of 96 various acylcarbenium ions in the gas phase and CH(3)CN were estimated by using the B3LYP/6-31+G(d)//B3LYP/6-31+G(d), B3LYP/6-311++G(2df,2p)//B3LYP/6-31+G(d), and BLYP/6-311++G(2df,2p)//B3LYP/6-31+G(d) methods for the first time. The results show that the combination of the BLYP/6-311++G(2df,2p)//B3LYP/6-31+G(d) method and IEFPCM could successfully predict the hydride affinities of arylcarbeniums in MeCN with a precision of about 3 kcal/mol. On the basis of the calculated results from the BLYP method, it can be found that the hydride affinity scale of the 96 arylcarbeniums in MeCN ranges from -130.76 kcal/mol for NO(2)-PhCH(+)-CN to -63.02 kcal/mol for p-(Me)(2)N-PhCH(+)-N(Me)(2), suggesting most of the arylcarbeniums are good hydride acceptors. Examination of the effect of the number of phenyl rings attached to the carbeniums on the hydride affinities shows that the increase of the hydride affinities takes place linearly with increasing number of benzene rings in the arylcarbeniums. Analyzing the effect of the substituents on the hydride affinities of arylcarbeniums indicates that electron-donating groups decrease the hydride affinities and electron-withdrawing groups show the opposite effect. The hydride affinities of arylcarbeniums are linearly dependent on the sum of the Hammett substituent parameters σ(p)(+). Inspection of the correlation of the solution-phase hydride affinities with gas-phase hydride affinities and aqueous-phase pK(R)(+) values reveals a remarkably good correspondence of ΔG(H(-)A)(R(+)) with both the gas-phase relative hydride affinities only if the α substituents X have no large electron-donating or -withdrawing properties and the pK(R)(+) values even though the media are dramatically different. The solution-phase hydride affinities also have a linear relationship with the electrophilicity parameter E, and this dependence can

Thermomechanics of candidate coatings for advanced gas reactor fuels

NASA Astrophysics Data System (ADS)

Nosek, A.; Conzen, J.; Doescher, H.; Martin, C.; Blanchard, J.

2007-09-01

Candidate fuel/coating combinations for an advanced, coated-fuel particle for a gas-cooled fast reactor (GFR) have been evaluated. These all-ceramic fuel forms consist of a fuel kernel made of UC or UN, surrounded with two shells (a buffer and a coating) made of TiC, SiC, ZrC, TiN, or ZrN. These carbides and nitrides are analyzed with finite element models to determine the stresses produced in the micro fuel particles from differential thermal expansion, fission gas release, swelling, and creep during particle fabrication and reactor operation. This study will help determine the feasibility of different fuel and coating combinations and identify the critical loads. The analysis shows that differential thermal expansion of the fuel and coating dictate the amount of stress for changing temperatures (such as during fabrication), and that the coating creep is able to mitigate an otherwise overwhelming amount of stress from fuel swelling. Because fracture is a likely mode of failure, a fracture mechanics study is also included to identify the relative likelihood of catastrophic fracture of the coating and resulting gas release. Overall, the analysis predicts that UN/ZrC is the best thermomechanical fuel/coating combination for mitigating the stress within the new fuel particle, but UN/TiN and UN/ZrN could also be strong candidates if their unknown creep rates are sufficiently large.

Structural and chemical degradation mechanisms of pure YSZ and its components ZrO2 and Y2O3 in carbon-rich fuel gases.

PubMed

Köck, Eva-Maria; Kogler, Michaela; Götsch, Thomas; Klötzer, Bernhard; Penner, Simon

2016-05-25

Structural and chemical degradation mechanisms of metal-free yttria stabilized zirconia (YSZ-8, 8 mol% Y2O3 in ZrO2) in comparison to its pure oxidic components ZrO2 and Y2O3 have been studied in carbon-rich fuel gases with respect to coking/graphitization and (oxy)carbide formation. By combining operando electrochemical impedance spectroscopy (EIS), operando Fourier-transform infrared (FT-IR) spectroscopy and X-ray photoelectron spectroscopy (XPS), the removal and suppression of CH4- and CO-induced carbon deposits and of those generated in more realistic fuel gas mixtures (syngas, mixtures of CH4 or CO with CO2 and H2O) was examined under SOFC-relevant conditions up to 1273 K and ambient pressures. Surface-near carbidization is a major problem already on the "isolated" (i.e. Nickel-free) cermet components, leading to irreversible changes of the conduction properties. Graphitic carbon deposition takes place already on the "isolated" oxides under sufficiently fuel-rich conditions, most pronounced in the pure gases CH4 and CO, but also significantly in fuel gas mixtures containing H2O and CO2. For YSZ, a comparative quantification of the total amount of deposited carbon in all gases and mixtures is provided and thus yields favorable and detrimental experimental approaches to suppress the carbon formation. In addition, the effectivity and reversibility of removal of the coke/graphite layers was comparably studied in the pure oxidants O2, CO2 and H2O and their effective contribution upon addition to the pure fuel gases CO and CH4 verified.

Hydride heat pump with heat regenerator

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor)

1991-01-01

A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

Porous metal hydride composite and preparation and uses thereof

DOEpatents

Steyert, W.A.; Olsen, C.E.

1980-03-12

A composite formed from large pieces of aggregate formed from (1) metal hydride (or hydride-former) powder and (2) either metal powder or plastic powder or both is prepared. The composite has large macroscopic interconnected pores (much larger than the sizes of the powders which are used) and will have a very fast heat transfer rate and low windage loss. It will be useful, for example, in heat engines, hydrogen storage devices, and refrigerator components which depend for their utility upon both a fast rate of hydriding and dehydriding. Additionally, a method of preparing the composite and a method of increasing the rates of hydriding and dehydriding of metal hydrides are also given.

Porous metal hydride composite and preparation and uses thereof

DOEpatents

Steyert, William A.; Olsen, Clayton E.

1982-01-01

A composite formed from large pieces of aggregate formed from (1) metal hydride (or hydride-former) powder and (2) either metal powder or plastic powder or both is prepared. The composite has large macroscopic interconnected pores (much larger than the sizes of the powders which are used) and will have a very fast heat transfer rate and low windage loss. It will be useful, for example, in heat engines, hydrogen storage devices, and refrigerator components which depend for their utility upon both a fast rate of hydriding and dehydriding. Additionally, a method of preparing the composite and a method of increasing the rates of hydriding and dehydriding of metal hydrides are also given.

Composite Materials for Hazard Mitigation of Reactive Metal Hydrides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pratt, Joseph William; Cordaro, Joseph Gabriel; Sartor, George B.

2012-02-01

In an attempt to mitigate the hazards associated with storing large quantities of reactive metal hydrides, polymer composite materials were synthesized and tested under simulated usage and accident conditions. The composites were made by polymerizing vinyl monomers using free-radical polymerization chemistry, in the presence of the metal hydride. Composites with vinyl-containing siloxane oligomers were also polymerized with and without added styrene and divinyl benzene. Hydrogen capacity measurements revealed that addition of the polymer to the metal hydride reduced the inherent hydrogen storage capacity of the material. The composites were found to be initially effective at reducing the amount of heatmore » released during oxidation. However, upon cycling the composites, the mitigating behavior was lost. While the polymer composites we investigated have mitigating potential and are physically robust, they undergo a chemical change upon cycling that makes them subsequently ineffective at mitigating heat release upon oxidation of the metal hydride. Acknowledgements The authors would like to thank the following people who participated in this project: Ned Stetson (U.S. Department of Energy) for sponsorship and support of the project. Ken Stewart (Sandia) for building the flow-through calorimeter and cycling test stations. Isidro Ruvalcaba, Jr. (Sandia) for qualitative experiments on the interaction of sodium alanate with water. Terry Johnson (Sandia) for sharing his expertise and knowledge of metal hydrides, and sodium alanate in particular. Marcina Moreno (Sandia) for programmatic assistance. John Khalil (United Technologies Research Corp) for insight into the hazards of reactive metal hydrides and real-world accident scenario experiments. Summary In an attempt to mitigate and/or manage hazards associated with storing bulk quantities of reactive metal hydrides, polymer composite materials (a mixture of a mitigating polymer and a metal hydride) were synthesized and

Storing hydrogen in the form of light alloy hydrides

NASA Technical Reports Server (NTRS)

Freund, E.; Gillerm, C.

1981-01-01

Different hydrides are investigated to find a system with a sufficiently high storage density (at least 3%). The formation of hydrides with light alloys is examined. Reaction kinetics for hydride formation were defined and applied to the systems Mg-Al-H, Mg-Al-Cu-H, Ti-Al-H, Ti-Al-Cu-H, and Ti-Al-Ni-H. Results indicate that the addition of Al destabilizes MgH2 and TiH2 hydrides while having only a limited effect on the storage density.

1. VIEW OF A PORTION OF THE HYDRIDE PROCESSING LABORATORY. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. VIEW OF A PORTION OF THE HYDRIDE PROCESSING LABORATORY. OPERATIONS IN THE GLOVE BOX IN THE BACKGROUND OF THE PHOTOGRAPH INCLUDED HYDRIDING OF PLUTONIUM AND HYDRIDE SEPARATION. IN THE FOREGROUND, THE VACUUM MONITOR CONTROL PANEL MEASURED TEMPERATURES WITHIN THE GLOVEBOX. THE CENTER CONTROL PANEL REGULATED THE FURNACE INSIDE THE GLOVE BOX USED IN THE HYDRIDING PROCESSES. THIS EQUIPMENT WAS ESSENTIAL TO THE HYDRIDING PROCESS, AS WELL AS OTHER GLOVE BOX OPERATIONS. - Rocky Flats Plant, Plutonium Laboratory, North-central section of industrial area at 79 Drive, Golden, Jefferson County, CO

Atomistic Simulation of High-Density Uranium Fuels

DOE PAGES

Garcés, Jorge Eduardo; Bozzolo, Guillermo

2011-01-01

We apply an atomistic modeling approach to deal with interfacial phenomena in high-density uranium fuels. The effects of Si, as additive to Al or as U-Mo-particles coating, on the behavior of the Al/U-Mo interface is modeled by using the Bozzolo-Ferrante-Smith (BFS) method for alloys. The basic experimental features characterizing the real system are identified, via simulations and atom-by-atom analysis. These include (1) the trend indicating formation of interfacial compounds, (2) much reduced diffusion of Al into U-Mo solid solution due to the high Si concentration, (3) Si depletion in the Al matrix, (4) an unexpected interaction between Mo and Simore » which inhibits Si diffusion to deeper layers in the U-Mo solid solution, and (5) the minimum amount of Si needed to perform as an effective diffusion barrier. Simulation results related to alternatives to Si dispersed in the Al matrix, such as the use of C coating of U-Mo particles or Zr instead of the Al matrix, are also shown. Recent experimental results confirmed early theoretical proposals, along the lines of the results reported in this work, showing that atomistic computational modeling could become a valuable tool to aid the experimental work in the development of nuclear fuels.« less

Investigation of ZrO x /ZrC-ZrN/Zr thin-film structural evolution and their degradation using X-ray diffraction and Raman spectrometry

NASA Astrophysics Data System (ADS)

Usmani, B.; Vijay, V.; Chhibber, R.; Dixit, A.

2016-11-01

The thin-film structures of DC/FR magnetron-sputtered ZrO x /ZrC-ZrN/Zr tandem solar-selective coatings are investigated using X-ray diffraction and room-temperature Raman spectroscopic measurements. These studies suggest that the major contribution is coming from h-ZrN0.28, c-ZrC, h-Zr3C2 crystallographic phases in ZrN-ZrC absorber layer, in conjunction with mixed ZrO x crystallographic phases. The change in structure for thermally annealed samples has been examined and observed that cubic and hexagonal ZrO x phase converted partially into tetragonal and monoclinic ZrO x phases, whereas hexagonal and cubic ZrN phases, from absorber layer, have not been observed for these thermally treated samples in air. These studies suggest that thermal treatment may lead to the loss of ZrN phase in absorber, degrading the thermal response for the desired wavelength range in open ambient conditions in contrast to vacuum conditions.

Bipolar Nickel-Metal Hydride Battery Being Developed

NASA Technical Reports Server (NTRS)

Manzo, Michelle A.

1998-01-01

The NASA Lewis Research Center has contracted with Electro Energy, Inc., to develop a bipolar nickel-metal hydride battery design for energy storage on low-Earth-orbit satellites. The objective of the bipolar nickel-metal hydride battery development program is to approach advanced battery development from a systems level while incorporating technology advances from the lightweight nickel electrode field, hydride development, and design developments from nickel-hydrogen systems. This will result in a low-volume, simplified, less-expensive battery system that is ideal for small spacecraft applications. The goals of the program are to develop a 1-kilowatt, 28-volt (V), bipolar nickel-metal hydride battery with a specific energy of 100 watt-hours per kilogram (W-hr/kg), an energy density of 250 W-hr/liter and a 5-year life in low Earth orbit at 40-percent depth-of-discharge.

Amorphization of the interaction products in U-Mo/Al dispersion fuel during irradiation

NASA Astrophysics Data System (ADS)

Ryu, Ho Jin; Kim, Yeon Soo; Hofman, G. L.

2009-04-01

The microstructures of the product resulting from interaction between U-Mo fuel particles and the Al matrix in U-Mo/Al dispersion fuel are discussed. We analyzed the available characterization results for the Al matrix dispersion fuels from both the out-of-pile and in-pile tests and examined the difference between these results. The morphology of pores that form in the interaction products during irradiation is similar to the porosity previously observed in irradiation-induced amorphized uranium compounds. The available diffraction studies for the interaction products formed in both the out-of-pile and in-pile tests are analyzed. We have concluded that the interaction products in the U-Mo/Al dispersion fuel are formed as an amorphous state or become amorphous during irradiation, depending on the irradiation conditions.

Fuel Cell Buses in U.S. Transit Fleets: Current Status 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eudy, Leslie; Post, Matthew; Jeffers, Matthew

This report, published annually, summarizes the progress of fuel cell electric bus development in the United States and discusses the achievements and challenges of introducing fuel cell propulsion in transit. The report provides a summary of results from evaluations performed by the National Renewable Energy Laboratory. Funding for this effort is provided by the U.S. Department of Energy's Fuel Cell Technologies Office within the Office of Energy Efficiency and Renewable Energy and by the U.S. Department of Transportation's Federal Transit Administration. The 2016 summary results primarily focus on the most recent year for each demonstration, from August 2015 through Julymore » 2016. The results for these buses account for more than 550,000 miles traveled and 59,500 hours of fuel cell power system operation. The primary results presented in the report are from three demonstrations of two different fuel-cell-dominant bus designs: Zero Emission Bay Area Demonstration Group led by Alameda-Contra Costa Transit District (AC Transit) in California; American Fuel Cell Bus Project at SunLine Transit Agency in California; and American Fuel Cell Bus Project at the University of California at Irvine.« less

In situ hydride formation in titanium during focused ion milling.

PubMed

Ding, Rengen; Jones, Ian P

2011-01-01

It is well known that titanium and its alloys are sensitive to electrolytes and thus hydrides are commonly observed in electropolished foils. In this study, focused ion beam (FIB) milling was used to prepare thin foils of titanium and its alloys for transmission electron microscopy. The results show the following: (i) titanium hydrides were observed in pure titanium, (ii) the preparation of a bulk sample in water or acid solution resulted in the formation of more hydrides and (iii) FIB milling aids the precipitation of hydrides, but there were never any hydrides in Ti64 and Ti5553.

Variation of Nb-Ta, Zr-Hf, Th-U and K-Cs in two diabase-granophyre suites

USGS Publications Warehouse

Gottfried, D.; Greenland, L.P.; Campbell, E.Y.

1968-01-01

Concentrations of Nb, Ta, Zr, Hf, Th, U and Cs have been determined in samples of igneous rocks representing the diabase-granophyre suites from Dillsburg, Pennsylvania, and Great Lake, Tasmania. Niobium and tantalum have a three to fourfold increase with differentiation in each of the suites. The chilled margin of the Great Lake intrusion contains half the niobium and tantalum content (5.3 ppm and 0.4 ppm, respectively) of the chilled basalt from Dillsburg (10 ppm and 0.9 ppm, respectively). The twofold difference between the suites is correlated with differences in their titanium content. The average Nb Ta ratios for each suite are similar: 13.5 for the Great Lake suite, and 14.4 for the Dillsburg suite. The zirconium content of the two suites is essentially the same and increases from 50 to 60 ppm in the chilled margins to 240-300 ppm in the granophyres. Hafnium is low in the early formed rocks (0.5 -1.5 ppm and achieves a maximum in the granophyres (5-8 ppm). The Zr Hfratio decreases from 68 to 33 with progressive differentiation. In the Dillsburg suite thorium and uranium increase from 2.6 ppm and 0.6 ppm, respectively, in the chilled samples to 11.8 ppm and 3.1 ppm in the granophyres. The chilled margin of the Great Lake suite contains 3.2 ppm thorium and 9.8 ppm uranium; the granophyre contains 11.2 ppm thorium and 2.8 ppm uranium. The average Th U ratios of the Dillsburg and Great Lake suites are nearly the same-4.1 and 4.4, respectively. Within each suite the Th U ratio remains quite constant. Cesium and the K Cs ratio do not vary systematically in the Dillsburg suite possibly because of redistribution or loss of cesium by complex geologic processes. Except for the chilled margin of the Great Lake suite, the variation of Cs and the K Cs ratio are in accord with theoretical considerations. Cesium increases from about 0.6 ppm in the lower zone to 3.5 ppm in the granophyre; the K Cs ratio varies from 10 ?? 103 in the lower zone to 6 ?? 103 in the granophyre. A

Irradiation induced structural change in Mo 2Zr intermetallic phase

DOE PAGES

Gan, J.; Keiser, Jr., D. D.; Miller, B. D.; ...

2016-05-14

The Mo 2Zr phase has been identified as a major interaction product at the interface of U-10Mo and Zr. Transmission electron microscopy in-situ irradiation with Kr ions at 200 °C with doses up to 2.0E+16 ions/cm 2 was carried out to investigate the radiation stability of the Mo 2Zr. The Mo 2Zr undergoes a radiation-induced structural change, from a large cubic (cF24) to a small cubic (cI2), along with an estimated 11.2% volume contraction without changing its composition. The structural change begins at irradiation dose below 1.0E+14 ions/cm 2. Furthermore, the transformed Mo 2Zr phase demonstrates exceptional radiation tolerance withmore » the development of dislocations without bubble formation.« less

A comparative study of the mechanical and thermal properties of defective ZrC, TiC and SiC.

PubMed

Jiang, M; Zheng, J W; Xiao, H Y; Liu, Z J; Zu, X T

2017-08-24

ZrC and TiC have been proposed to be alternatives to SiC as fuel-cladding and structural materials in nuclear reactors due to their strong radiation tolerance and high thermal conductivity at high temperatures. To unravel how the presence of defects affects the thermo-physical properties under irradiation, first-principles calculations based on density function theory were carried out to investigate the mechanical and thermal properties of defective ZrC, TiC and SiC. As compared with the defective SiC, the ZrC and TiC always exhibit larger bulk modulus, smaller changes in the Young's and shear moduli, as well as better ductility. The total thermal conductivity of ZrC and TiC are much larger than that of SiC, implying that under radiation environment the ZrC and TiC will exhibit superior heat conduction ability than the SiC. One disadvantage for ZrC and TiC is that their Debye temperatures are generally lower than that of SiC. These results suggest that further improving the Debye temperature of ZrC and TiC will be more beneficial for their applications as fuel-cladding and structural materials in nuclear reactors.

Accident-tolerant oxide fuel and cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, Robert D.

Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion.more » The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.« less

Analysis of fuel cycle strategies and U.S. transition scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wigeland, Roald; Taiwo, Temitope A.

2016-10-17

The nuclear fuel cycle Evaluation and Screening (E&S) study that was completed in October 2014 [1] enabled the identification of four fuel cycle groups that are considered most promising based on a set of nine evaluation criteria: (a) six benefit criteria of Nuclear Waste Management, Proliferation Risk, Nuclear Material Security Risk, Safety, Environmental Impact, Resource Utilization, and (b) three challenge criteria of Development and Deployment Risk, Institutional Issues, Financial Risk and Economics. The E&S study was conducted at a level of analysis that is "technology- neutral," that is, without consideration of specific technologies, but using the fundamental physics characteristics ofmore » each part of the fuel cycle. The study focused on the fuel cycle performance benefits at the fuel cycle equilibrium state, with only limited consideration of transition and deployment impacts. Common characteristics of the four most promising fuel cycle options include continuous recycle of all U/Pu or U/TRU, the use of fast-spectrum reactors, and no use of uranium enrichment once fuel cycle equilibrium has been established. The high-level wastes are mainly from processing of irradiated fuel, and there would be no disposal of any spent fuel. Building on the findings of the E&S study, additional studies have been conducted in the last two years following the information exchange meeting, the 13th IEMPT, which was held in Seoul, the Republic of Korea in 2014. Insights are presented from the recent studies on the benefits and challenges of recycling minor actinides, and transition considerations to some of the most promising fuel cycle options.« less

Titanite chronology, thermometry, and speedometry of ultrahigh-temperature (UHT) calc-silicates from south Madagascar: U-Pb dates, Zr temperatures, and lengthscales of trace-element diffusion

NASA Astrophysics Data System (ADS)

Holder, R. M.; Hacker, B. R.

2017-12-01

Calc-silicate rocks are often overlooked as sources of pressure-temperature-time data in granulite-UHT metamorphic terranes due to the strong dependence of calc-silicate mineral assemblages on complex fluid compositions and a lack of thermodynamic data on common high-temperature calc-silicate minerals such as scapolite. In the Ediacaran-Cambrian UHT rocks of southern Madagascar, clinopyroxene-scapolite-feldspar-quartz-zircon-titanite calc-silicate rocks are wide-spread. U-Pb dates of 540-520 Ma from unaltered portions of titanite correspond to cooling of the rocks through upper-amphibolite facies and indicate UHT metamorphism occurred before 540 Ma. Zr concentrations in these domains preserve growth temperatures of 900-950 °C, consistent with peak temperatures calculated by pseudosection modeling of nearby osumilite-bearing gneisses. Younger U-Pb dates (510-490 Ma) correspond to fluid-mediated Pb loss from titanite grains, which occurred below their diffusive Pb-closure temperature, along fractures. The extent of fluid alteration is seen clearly in back-scattered electron images and Zr-, Al-, Fe-, Ce-, and Nb-concentration maps. Laser-ablation depth profiling of idioblastic titanite grains shows preserved Pb diffusion profiles at grain rims, but there is no evidence for Zr diffusion, indicating that it was effectively immobile even at UHT.

Low density metal hydride foams

DOEpatents

Maienschein, Jon L.; Barry, Patrick E.

1991-01-01

Disclosed is a low density foam having a porosity of from 0 to 98% and a density less than about 0.67 gm/cc, prepared by heating a mixture of powered lithium hydride and beryllium hydride in an inert atmosphere at a temperature ranging from about 455 to about 490 K for a period of time sufficient to cause foaming of said mixture, and cooling the foam thus produced. Also disclosed is the process of making the foam.

Effect of High Si Content on U3Si2 Fuel Microstructure

NASA Astrophysics Data System (ADS)

Rosales, Jhonathan; van Rooyen, Isabella J.; Meher, Subhashish; Hoggan, Rita; Parga, Clemente; Harp, Jason

2018-02-01

The development of U3Si2 as an accident-tolerant nuclear fuel has gained research interest because of its promising high uranium density and improved thermal properties. In the present study, three samples of U3Si2 fuel with varying silicon content have been fabricated by a conventional powder metallurgical route. Microstructural characterization via scanning and transmission electron microscopy reveals the presence of other stoichiometry of uranium silicide such as USi and UO2 in both samples. The detailed phase analysis by x-ray diffraction shows the presence of secondary phases, such as USi, U3Si, and UO2. The samples with higher concentrations of silicon content of 7.5 wt.% display additional elemental Si. These samples also possess an increased amount of the USi phase as compared to that in the conventional sample with 7.3 wt.% silicon. The optimization of U3Si2 fuel performance through the understanding of the role of Si content on its microstructure has been discussed.

Method of making crack-free zirconium hydride

DOEpatents

Sullivan, Richard W.

1980-01-01

Crack-free hydrides of zirconium and zirconium-uranium alloys are produced by alloying the zirconium or zirconium-uranium alloy with beryllium, or nickel, or beryllium and scandium, or nickel and scandium, or beryllium and nickel, or beryllium, nickel and scandium and thereafter hydriding.

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

Thermodynamic and Kinetic Properties of Metal Hydrides from First-Principles Calculations

NASA Astrophysics Data System (ADS)

Michel, Kyle Jay

In an effort to minimize the worldwide dependence on fossil fuels, much research has focused on the development of hydrogen fuel cell vehicles. Among the many challenges currently facing the transition to such an alternative energy economy is the storage of hydrogen in an economical and practical way. One class of materials that has presented itself as a possible candidate is solid metal hydrides. These materials chemically bind hydrogen and on heating, release the gas which can then be used to generate power as needed for the vehicle. In order to meet guidelines that have been set for such a storage system, hydrogen must be released rapidly in a narrow temperature range of -40 to 80°C with all reactions being reversible. This sets both thermodynamic and kinetic requirements for the design of candidate metal hydrides. First-principles calculations are well-suited for the task of exploring reactions involving metal hydrides. Here, density-functional theory is used to calculate properties of these materials at the quantum mechanical level of accuracy. In particular, three systems have been investigated: 1. Li-Mg-N-H. Reactions between all known compounds in this system are systematically investigated in order to predict thermodynamically allowed reactions that release hydrogen. The properties of these reactions are compared to the requirements set for hydrogen storage systems. Additionally, ground-state structures are predicted for Li2Mg(NH)2 and Li 4Mg(NH)3. 2. Na-Al-H. The kinetics of mass transport during the (de)hydrogenation of the well-known metal hydride NaAlH4 are investigated. A model is developed to study the flux of native defects through phases involved in these reactions. Since it is also known that titanium is an effective catalyst for both dehydrogenation and rehydrogenation, the effect of Ti substitution in bulk lattices on the kinetics of mass transport is investigated. Results are compared to experiments in order to determine if mass transport

Zr/ZrC modified layer formed on AISI 440B stainless steel by plasma Zr-alloying

NASA Astrophysics Data System (ADS)

Shen, H. H.; Liu, L.; Liu, X. Z.; Guo, Q.; Meng, T. X.; Wang, Z. X.; Yang, H. J.; Liu, X. P.

2016-12-01

The surface Zr/ZrC gradient alloying layer was prepared by double glow plasma surface alloying technique to increase the surface hardness and wear resistance of AISI 440B stainless steel. The microstructure of the Zr/ZrC alloying layer formed at different alloying temperatures and times as well as its formation mechanism were discussed by using scanning electron microscopy, glow discharge optical emission spectrum, X-ray diffraction and X-ray photoelectron spectroscopy. The adhesive strength, hardness and tribological property of the Zr/ZrC alloying layer were also evaluated in the paper. The alloying surface consists of the Zr-top layer and ZrC-subsurface layer which adheres strongly to the AISI 440B steel substrate. The thickness of the Zr/ZrC alloying layer increases gradually from 16 μm to 23 μm with alloying temperature elevated from 900 °C to 1000 °C. With alloying time from 0.5 h to 4 h, the alloyed depth increases from 3 μm to 30 μm, and the ZrC-rich alloyed thickness vs time is basically parabola at temperature of 1000 °C. Both the hardness and wear resistance of the Zr/ZrC alloying layer obviously increase compared with untreated AISI 440B steel.

Oxidative desulfurization of benzothiophene and thiophene with WOx/ZrO2 catalysts: effect of calcination temperature of catalysts.

PubMed

Hasan, Zubair; Jeon, Jaewoo; Jhung, Sung Hwa

2012-02-29

Oxidative desulfurization (ODS) of model fuel containing benzothiophene (BT) or thiophene (Th) has been carried out with WO(x)/ZrO2 catalyst, which was calcined at various temperatures. Based on the conversion of BT in the model fuel, it can be shown that the optimum calcination temperature of WOx/ZrO2 catalyst is around 700 °C. The most active catalyst is composed of tetragonal zirconia (ZrO2) with well dispersed polyoxotungstate species and it is necessary to minimize the contents of the crystalline WO3 and monoclinic ZrO2 for a high BT conversion. The oxidation rate was interpreted with the first-order kinetics, and it demonstrated the importance of electron density since the kinetic constant for BT was higher than that for Th even though the BT is larger than Th in size. A WOx/ZrO2 catalyst, treated suitably, can be used as a reusable active catalyst in the ODS. Copyright © 2012 Elsevier B.V. All rights reserved.

High H⁻ ionic conductivity in barium hydride.

PubMed

Verbraeken, Maarten C; Cheung, Chaksum; Suard, Emmanuelle; Irvine, John T S

2015-01-01

With hydrogen being seen as a key renewable energy vector, the search for materials exhibiting fast hydrogen transport becomes ever more important. Not only do hydrogen storage materials require high mobility of hydrogen in the solid state, but the efficiency of electrochemical devices is also largely determined by fast ionic transport. Although the heavy alkaline-earth hydrides are of limited interest for their hydrogen storage potential, owing to low gravimetric densities, their ionic nature may prove useful in new electrochemical applications, especially as an ionically conducting electrolyte material. Here we show that barium hydride shows fast pure ionic transport of hydride ions (H(-)) in the high-temperature, high-symmetry phase. Although some conductivity studies have been reported on related materials previously, the nature of the charge carriers has not been determined. BaH2 gives rise to hydride ion conductivity of 0.2 S cm(-1) at 630 °C. This is an order of magnitude larger than that of state-of-the-art proton-conducting perovskites or oxide ion conductors at this temperature. These results suggest that the alkaline-earth hydrides form an important new family of materials, with potential use in a number of applications, such as separation membranes, electrochemical reactors and so on.

Reactivity of yttrium carboxylates toward alkylaluminum hydrides.

PubMed

Schädle, Christoph; Fischbach, Andreas; Herdtweck, Eberhardt; Törnroos, Karl W; Anwander, Reiner

2013-11-25

Yttrocene-carboxylate complex [Cp*2Y(OOCAr(Me))] (Cp*=C5Me5, Ar(Me) =C6H2Me3-2,4,6) was synthesized as a spectroscopically versatile model system for investigating the reactivity of alkylaluminum hydrides towards rare-earth-metal carboxylates. Equimolar reactions with bis-neosilylaluminum hydride and dimethylaluminum hydride gave adduct complexes of the general formula [Cp*2Y(μ-OOCAr(Me))(μ-H)AlR2] (R=CH2SiMe3, Me). The use of an excess of the respective aluminum hydride led to the formation of product mixtures, from which the yttrium-aluminum-hydride complex [{Cp*2Y(μ-H)AlMe2(μ-H)AlMe2(μ-CH3)}2] could be isolated, which features a 12-membered-ring structure. The adduct complexes [Cp*2Y(μ-OOCAr(Me))(μ-H)AlR2] display identical (1)J(Y,H) coupling constants of 24.5 Hz for the bridging hydrido ligands and similar (89)Y NMR shifts of δ=-88.1 ppm (R=CH2SiMe3) and δ=-86.3 ppm (R=Me) in the (89)Y DEPT45 NMR experiments. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Magnetic properties and crystal structure of RENiA1 and UniA1 hydrides.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bordallo, H. N.; Drulis, H.; Havela, L.

1999-08-11

RENiAl (RE = rare-earth metal) and UNiAl compounds crystallizing in the hexagonal ZrNiAl-type structure (space group P{bar 6}2m) can absorb up to 2 and 3 hydrogen (deuterium) atoms per formula unit, respectively. Hydrogenation leads to a notable lattice expansion and modification of magnetic properties. However, the impact of hydrogenation on magnetism is the opposite for 4f- and 5f-materials: TN(T{sub c})is lowered in the case of rare-earth hydrides, while for UNiAlH(D){sub x} it increases by an order of magnitude. Here we present results of magnetic and structure studies performed of these compounds, focusing on the correlation between magnetic and structural variationsmore » and discussing possible reasons of the striking difference in effect of hydrogenation on rare-earth and actinide intermetallics.« less

Experimental investigation of paraffin-based fuels for hybrid rocket propulsion

NASA Astrophysics Data System (ADS)

Galfetti, L.; Merotto, L.; Boiocchi, M.; Maggi, F.; DeLuca, L. T.

2013-03-01

Solid fuels for hybrid rockets were characterized in the framework of a research project aimed to develop a new generation of solid fuels, combining at the same time good mechanical and ballistic properties. Original techniques were implemented in order to improve paraffin-based fuels. The first strengthening technique involves the use of a polyurethane foam (PUF); a second technique is based on thermoplastic polymers mixed at molecular level with the paraffin binder. A ballistic characterization of paraffin-based hybrid rocket solid fuels was performed, considering pure wax-based fuels and fuels doped with suitable metal additives. Nano-Al powders and metal hydrides (magnesium hydride (MgH2), lithium aluminum hydride (LiAlH4 )) were used as fillers in paraffin matrices. The results of this investigation show a strong correlation between the measured viscosity of the melted paraffin layer and the regression rate: a decrease of viscosity increases the regression rate. This trend is due to the increasing development of entrainment phenomena, which strongly increase the regression rate. Addition of LiAlH4 (mass fraction 10%) can further increase the regression rate up to 378% with respect to the pure HTPB regression rate, taken as baseline reference fuel. The highest regression rates were found for the Solid Wax (SW) composition, added with 5% MgH2 mass fraction; at 350 kg/(m2s) oxygen mass flux, the measured regression rate, averaged in space and time, was 2.5 mm/s, which is approximately five times higher than that of the pure HTPB composition. Compositions added with nanosized aluminum powders were compared with those added with MgH2, using gel or solid wax.

Synthesis, spectral, DFT modeling, cytotoxicity and microbial studies of novel Zr(IV), Ce(IV) and U(VI) piroxicam complexes

NASA Astrophysics Data System (ADS)

El-Shwiniy, Walaa H.; Zordok, Wael A.

2018-06-01

The Zr(IV), Ce(IV) and U(VI) piroxicam anti-inflammatory drug complexes were prepared and characterized using elemental analyses, conductance, IR, UV-Vis, magnetic moment, IHNMR and thermal analysis. The ratio of metal: Pir is found to be 1:2 in all complexes estimated by using molar ratio method. The conductance data reveal that Zr(IV) and U(VI) chelates are non-electrolytes except Ce(IV) complex is electrolyte. Infrared spectroscopic confirm that the Pir behaves as a bidentate ligand co-ordinated to the metal ions via the oxygen and nitrogen atoms of ν(Cdbnd O)carbonyl and ν(Cdbnd N)pyridyl, respectively. The kinetic parameters of thermogravimetric and its differential, such as activation energy, entropy of activation, enthalpy of activation, and Gibbs free energy evaluated using Coats-Redfern and Horowitz-Metzger equations for Pir and complexes. The geometry of the piroxicam drug in the Free State differs significantly from that in the metal complex. In the time of metal ion-drug bond formation the drug switches-on from the closed structure (equilibrium geometry) to the open one. The antimicrobial tests were assessed towards some types of bacteria and fungi. The in vitro cell cytotoxicity of the complexes in comparison with Pir against colon carcinoma (HCT-116) cell line was measured. Optimized geometrical structure of piroxicam ligand by using DFT calculations.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1999-01-01

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Promoting the ambient-condition stability of Zr-doped barium cerate: Toward robust solid oxide fuel cells and hydrogen separation in syngas

NASA Astrophysics Data System (ADS)

Yang, Ying; Zeng, Yimin; Amirkhiz, Babak S.; Luo, Jing-Li; Yan, Ning

2018-02-01

Increasing the stability of perovskite proton conductor against atmospheric CO2 and moisture attack at ambient conditions might be equally important as that at the elevated service temperatures. It can ease the transportation and storage of materials, potentially reducing the maintenance cost of the integral devices. In this work, we initially examined the surface degradation behaviors of various Zr-doped barium cerates (BaCe0.7Zr0.1Y0.1Me0.1O3) using XRD, SEM, STEM and electron energy loss spectroscopy. Though that the typical lanthanide (Y, Yb and Gd) and In incorporated Zr-doped cerates well resisted CO2-induced carbonation in air at elevated temperatures, they were unfortunately vulnerable at ambient conditions, suffering slow decompositions at the surface. Conversely, Sn doped samples (BCZYSn) were robust at both conditions yet showed high protonic conductivity. Thanks to that, the anode supported solid oxide fuel cells equipped with BCZYSn electrolyte delivered a maximum power density of 387 mW cm-2 at 600 °C in simulated coal-derived syngas. In the hydrogen permeation test using BCZYSn based membrane, the H2 flux reached 0.11 mL cm-2 min-1 at 850 °C when syngas was the feedstock. Both devices demonstrated excellent stability in the presence of CO2 in the syngas.

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sears, D.F.; Primeau, M.F.; Buchanan, C.

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinationsmore » showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.« less

Phenylsilane as a safe, versatile alternative to hydrogen for the synthesis of actinide hydrides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pagano, Justin K.; Dorhout, Jacquelyn M.; Waterman, Rory

2015-10-22

The thorium and uranium dihydride dimer complexes [(C 5Me 5) 2An(H)(μ-H)] 2 (An = Th, U) have been easily prepared using phenylsilane, which is an efficient and safer alternative to hydrogen gas. We demonstrated the synthetic utility of this new hydriding method by the preparation of a variety of organometallic complexes, including, for the first time, (C 5Me 5) 2U(SMe) 2, (C 5Me 5) 2Th(C 4Ph 4), (C 5Me 5) 2U(C 4Ph 4), (C 5Me 5) 2ThS 5, and (C 5Me 5) 2U(bipy) using [(C 5Me 5) 2An(H)(μ-H)] 2 (An = Th, U) as multi-electron reductants.

Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Suman, Siddharth; Khan, Mohd Kaleem; Pathak, Manabendra; Singh, R. N.

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.

Structural and kinetic studies of metal hydride hydrogen storage materials using thin film deposition and characterization techniques

NASA Astrophysics Data System (ADS)

Kelly, Stephen Thomas

Hydrogen makes an attractive energy carrier for many reasons. It is an abundant chemical fuel that can be produced from a wide variety of sources and stored for very long periods of time. When used in a fuel cell, hydrogen emits only water at the point of use, making it very attractive for mobile applications such as in an automobile. Metal hydrides are promising candidates for on-board reversible hydrogen storage in mobile applications due to their very high volumetric storage capacities---in most cases exceeding even that of liquid hydrogen. The United States Department of Energy (DOE) has set fuel system targets for an automotive hydrogen storage system, but as of yet no single material meets all the requirements. In particular, slow reaction kinetics and/or inappropriate thermodynamics plague many metal hydride hydrogen storage materials. In order to engineer a practical material that meets the DOE targets, we need a detailed understanding of the kinetic and thermodynamic properties of these materials during the phase change. In this work I employed sputter deposited thin films as a platform to study materials with highly controlled chemistry, microstructure and catalyst placement using thin film characterization techniques such as in situ x-ray diffraction (XRD) and neutron reflectivity. I observed kinetic limitations in the destabilized Mg2Si system due to the slow diffusion of the host Mg and Si atoms while forming separate MgH2 and Si phases. Conversely, I observed that the presence of Al in the Mg/Al system inhibits hydrogen diffusion while the host Mg and Al atoms interdiffuse readily, allowing the material to fall into a kinetic and/or thermodynamic trap by forming intermetallic compounds such as Mg17Al 12. By using in situ XRD to analyze epitaxial Mg films grown on (001) oriented Al2O3 substrates I observed hydride growth consistent with a model of a planar hydride layer growing into an existing metal layer. Subsequent film cycling changes the hydrogen

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

Microstructural Modeling of Dynamic Intergranular and Transgranular Fracture Modes in Zircaloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohammed, I.; Zikry, M.A.; Ziaei, S.

2017-04-01

In this time period, we have continued to focus on (i) refining the thermo-mechanical fracture model for zirconium (Zr) alloys subjected to large deformations and high temperatures that accounts for the cracking of ZrH and ZrH2 hydrides, (ii) formulating a framework to account intergranular fracture due to iodine diffusion and pit formation in grain-boundaries (GBs). Our future objectives are focused on extending to a combined population of ZrH and ZrH2 populations and understanding how thermo-mechanical behavior affects hydride reorientation and cracking. We will also refine the intergranular failure mechanisms for grain boundaries with pits.

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

Uranium Hydride Nucleation and Growth Model FY'16 ESC Annual Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, Mary Ann; Richards, Andrew Walter; Holby, Edward F.

2016-12-20

Uranium hydride corrosion is of great interest to the nuclear industry. Uranium reacts with water and/or hydrogen to form uranium hydride which adversely affects material performance. Hydride nucleation is influenced by thermal history, mechanical defects, oxide thickness, and chemical defects. Information has been gathered from past hydride experiments to formulate a uranium hydride model to be used in a Canned Subassembly (CSA) lifetime prediction model. This multi-scale computer modeling effort started in FY’13, and the fourth generation model is now complete. Additional high-resolution experiments will be run to further test the model.

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın

2011-07-01

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

Separation of the rare-earth fission product poisons from spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christian, Jerry D.; Sterbentz, James W.

A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2more » in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.« less

Inhibitive effect of Pt on Pd-hydride formation of Pd@Pt core-shell electrocatalysts: An in situ EXAFS and XRD study

DOE PAGES

Wise, Anna M.; Richardson, Peter W.; Price, Stephen W. T.; ...

2017-12-27

In situ EXAFS and XRD have been used to study the electrochemical formation of hydride phases, H abs, in 0.5 M H 2SO 4 for a Pd/C catalyst and a series of Pd@Pt core-shell catalysts with varying Pt shell thickness, from 0.5 to 4 monolayers. Based on the XRD data a 3% lattice expansion is observed for the Pd/C core catalyst upon hydride formation at 0.0 V. In contrast, the expansion was ≤0.6% for all of the core-shell catalysts. The limited extent of the lattice expansion observed suggests that hydride formation, which may occur during periodic active surface area measurementsmore » conducting during accelerated aging tests or driven by H 2 crossover in PEM fuel cells, is unlikely to contribute significantly to the degradation of Pd@Pt core-shell electrocatalysts in contrast to the effects of oxide formation.« less

Inhibitive effect of Pt on Pd-hydride formation of Pd@Pt core-shell electrocatalysts: An in situ EXAFS and XRD study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wise, Anna M.; Richardson, Peter W.; Price, Stephen W. T.

In situ EXAFS and XRD have been used to study the electrochemical formation of hydride phases, H abs, in 0.5 M H 2SO 4 for a Pd/C catalyst and a series of Pd@Pt core-shell catalysts with varying Pt shell thickness, from 0.5 to 4 monolayers. Based on the XRD data a 3% lattice expansion is observed for the Pd/C core catalyst upon hydride formation at 0.0 V. In contrast, the expansion was ≤0.6% for all of the core-shell catalysts. The limited extent of the lattice expansion observed suggests that hydride formation, which may occur during periodic active surface area measurementsmore » conducting during accelerated aging tests or driven by H 2 crossover in PEM fuel cells, is unlikely to contribute significantly to the degradation of Pd@Pt core-shell electrocatalysts in contrast to the effects of oxide formation.« less

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hermann, S.D.; Gese, N.J.; Wurth, L.A.

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide.more » In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.« less

Effect of indium addition in U-Zr metallic fuel on lanthanide migration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Yeon Soo; Wiencek, T.; O'Hare, E.

Advanced fast reactor concepts to achieve ultra-high burnup (~50%) require prevention of fuel-cladding chemical interaction (FCCI). Fission product lanthanide accumulation at high burnup is substantial and significantly contributes to FCCI upon migration to the cladding interface. Diffusion barriers are typically used to prevent interaction of the lanthanides with the cladding. A more active method has been proposed which immobilizes the lanthanides through formation of stable compounds with an additive. Theoretical analysis showed that indium, thallium, and antimony are good candidates. Indium was the strongest candidate because of its low reactivity with iron-based cladding alloys. Characterization of the as-fabricated alloys wasmore » performed to determine the effectiveness of the indium addition in forming compounds with lanthanides, represented by cerium. Tests to examine how effectively the dopant prevents lanthanide migration under a thermal gradient were also performed. The results showed that indium effectively prevented cerium migration.« less

Advanced Catalysts for Fuel Cells

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R.; Whitacre, Jay; Valdez, T. I.

2006-01-01

This viewgraph presentation reviews the development of catalyst for Fuel Cells. The objectives of the project are to reduce the cost of stack components and reduce the amount of precious metal used in fuel cell construction. A rapid combinatorial screening technique based on multi-electrode thin film array has been developed and validated for identifying catalysts for oxygen reduction; focus shifted from methanol oxidation in FY05 to oxygen reduction in FY06. Multi-electrode arrays of thin film catalysts of Pt-Ni and Pt-Ni-Zr have been deposited. Pt-Ni and have been characterized electrochemically and structurally. Pt-Ni-Zr and Pt-Ni films show higher current density and onset potential compared to Pt. Electrocatalytic activity and onset potential are found to be strong function of the lattice constant. Thin film Pt(59)Ni(39)Zr(2) can provide 10 times the current density of thin film Pt. Thin film Pt(59)Ni(39)Zr(2) also shows 65mV higher onset potential than Pt.

Thermal properties of U-7Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong-Tak; Yang, Jae Ho; Oh, Jang Soo; Won, Ju-Jin; Sohn, Dong-Seong

2017-12-01

The thermal diffusivity and heat capacity of U-7Mo/Al and U-7Mo/Al-5Si as functions of U-Mo fuel volume fraction and temperature were measured. The density of the sample was measured at room temperature and estimated using thermal expansion data at elevated temperatures. Using the measured data, the thermal conductivity was obtained as a function of U-Mo volume fraction and temperature. The thermal conductivity of U-7Mo/Al-5Si was found to be lower than that of U-7Mo/Al because of the Si addition to the Al. Due to a lower porosity and reduced interaction between U-Mo and Al in the sample, the thermal conductivity data reported in the present study were higher than those in the literature. The present data were found to be in agreement with the predictions of theoretical models.

Numerical comparison of hydrogen desorption behaviors of metal hydride beds based on uranium and on zirconium-cobalt

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kyoung, S.; Yoo, H.; Ju, H.

2015-03-15

In this paper, the hydrogen delivery capabilities of uranium (U) and zirconium-cobalt (ZrCo) are compared quantitatively in order to find the optimum getter materials for tritium storage. A three-dimensional hydrogen desorption model is applied to two identically designed cylindrical beds with the different materials, and hydrogen desorption simulations are then conducted. The simulation results show superior hydrogen delivery performance and easier thermal management capability for the U bed. This detailed analysis of the hydrogen desorption behaviors of beds with U and ZrCo will help to identify the optimal bed material, bed design, and operating conditions for the storage and deliverymore » system in ITER. (authors)« less

Development of Metallic Fuels for Actinide Transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, Steven Lowe; Fielding, Randall Sidney; Benson, Michael Timothy

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimizedmore » for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

The free-energy barrier to hydride transfer across a dipalladium complex

DOE PAGES

Ramirez-Cuesta, Anibal J.

2015-01-01

We use density-functional theory molecular dynamics (DFT-MD) simulations to determine the hydride transfer coordinate between palladium centres of the crystallographically observed terminal hydride locations, Pd-Pd-H, originally postulated for the solution dynamics of the complex bis-NHC dipalladium hydride [{(MesIm)(2)CH2}(2)Pd2H][PF6], and then calculate the free-energy along this coordinate. We estimate the transfer barrier-height to be about 20 kcal mol(-1) with a hydride transfer rate in the order of seconds at room temperature. We validate our DFT-MD modelling using inelastic neutron scattering which reveals anharmonicity of the hydride environment that is so pronounced that there is complete failure of the harmonic model formore » the hydride ligand. The simulations are extended to high temperature to bring the H-transfer to a rate that is accessible to the simulation technique.« less

Destabilisation of complex hydrides through size effects.

PubMed

Christian, Meganne; Aguey-Zinsou, Kondo-Francois

2010-12-01

Nanoparticles of NaAlH4, LiAlH4 and LiBH4 were prepared by encapsulating their respective hydrides within carbon nanotubes by a wet chemical approach. The resulting confinement had a profound effect on the overall hydrogen storage properties of these hydrides, with NaAlH4 and LiAlH4 releasing hydrogen from room temperature, for example.

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

DOE PAGES

Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

2016-05-01

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

AIR PASSIVATION OF METAL HYDRIDE BEDS FOR WASTE DISPOSAL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, J; R. H. Hsu, R

2007-07-02

Metal hydride beds offer compact, safe storage of tritium. After metal hydride beds have reached the end of their useful life, the beds will replaced with new beds and the old beds prepared for disposal. One acceptance criteria for hydride bed waste disposal is that the material inside the bed not be pyrophoric. To determine the pyrophoric nature of spent metal hydride beds, controlled air ingress tests were performed. A simple gas handling manifold fitted with pressure transducers and a calibrated volume were used to introduce controlled quantities of air into a metal hydride bed and the bed temperature risemore » monitored for reactivity with the air. A desorbed, 4.4 kg titanium prototype hydride storage vessel (HSV) produced a 4.4 C internal temperature rise upon the first air exposure cycle and a 0.1 C temperature rise upon a second air exposure. A total of 346 scc air was consumed by the bed (0.08 scc per gram Ti). A desorbed, 9.66 kg LaNi{sub 4.25}Al{sub 0.75} prototype storage bed experienced larger temperature rises over successive cycles of air ingress and evacuation. The cycles were performed over a period of days with the bed effectively passivated after the 12th cycle. Nine to ten STP-L of air reacted with the bed producing both oxidized metal and water.« less

In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Silva, G W Chinthaka M; Kiggans, Jim

2013-01-01

The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels was examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was least substantial while PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantlymore » faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.« less

High energy density battery based on complex hydrides

DOEpatents

Zidan, Ragaiy

2016-04-26

A battery and process of operating a battery system is provided using high hydrogen capacity complex hydrides in an organic non-aqueous solvent that allows the transport of hydride ions such as AlH.sub.4.sup.- and metal ions during respective discharging and charging steps.

Hydrogen and dihydrogen bonding of transition metal hydrides

NASA Astrophysics Data System (ADS)

Jacobsen, Heiko

2008-04-01

Intermolecular interactions between a prototypical transition metal hydride WH(CO) 2NO(PH 3) 2 and a small proton donor H 2O have been studied using DFT methodology. The hydride, nitrosyl and carbonyl ligand have been considered as site of protonation. Further, DFT-D calculations in which empirical corrections for the dispersion energy are included, have been carried out. A variety of pure and hybrid density functionals (BP86, PW91, PBE, BLYP, OLYP, B3LYP, B1PW91, PBE0, X3LYP) have been considered, and our calculations indicate the PBE functional and its hybrid variation are well suited for the calculation of transition metal hydride hydrogen and dihydrogen bonding. Dispersive interactions make up for a sizeable portion of the intermolecular interaction, and amount to 20-30% of the bond energy and to 30-40% of the bond enthalpy. An energy decomposition analysis reveals that the H⋯H bond of transition metal hydrides contains both covalent and electrostatic contributions.

Regenerative Hydride Heat Pump

NASA Technical Reports Server (NTRS)

Jones, Jack A.

1992-01-01

Hydride heat pump features regenerative heating and single circulation loop. Counterflow heat exchangers accommodate different temperatures of FeTi and LaNi4.7Al0.3 subloops. Heating scheme increases efficiency.

On the Chemistry of Hydrides of N Atoms and O+ Ions

NASA Astrophysics Data System (ADS)

Awad, Zainab; Viti, Serena; Williams, David A.

2016-08-01

Previous work by various authors has suggested that the detection by Herschel/HIFI of nitrogen hydrides along the low-density lines of sight toward G10.6-0.4 (W31C) cannot be accounted for by gas-phase chemical models. In this paper we investigate the role of surface reactions on dust grains in diffuse regions, and we find that formation of the hydrides by surface reactions on dust grains with efficiency comparable to that for H2 formation reconciles models with observations of nitrogen hydrides. However, similar surface reactions do not contribute significantly to the hydrides of O+ ions detected by Herschel/HIFI that are present along many sight lines in the Galaxy. The O+ hydrides can be accounted for by conventional gas-phase chemistry either in diffuse clouds of very low density with normal cosmic-ray fluxes or in somewhat denser diffuse clouds with high cosmic-ray fluxes. Hydride chemistry in dense dark clouds appears to be dominated by gas-phase ion-molecule reactions.

Enhancement of Oxidative Desulfurization Performance over UiO-66(Zr) by Titanium Ion Exchange.

PubMed

Ye, Gan; Qi, Hui; Li, Xiaolin; Leng, Kunyue; Sun, Yinyong; Xu, Wei

2017-07-19

Oxidative desulfurization is considered to be one of the most promising methods for producing ultra-low-sulfur fuels because it can effectively remove refractory sulfur-containing aromatic compounds under mild conditions. In this work, the oxidative desulfurization performance over UiO-66(Zr) is greatly enhanced by Ti ion exchange. This strategy is not only efficient for UiO-66(Zr) with crystal defects but also for UiO-66(Zr) with high crystallinity. In particular, the performance of UiO-66(Zr) with high crystallinity in the oxidative desulfurization of dibenzothiophene can be improved more than 11-fold, which can be mainly attributed to the introduction of active Ti sites. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Low-valent group 14 element hydride chemistry: towards catalysis.

PubMed

Hadlington, Terrance J; Driess, Matthias; Jones, Cameron

2018-06-05

The chemistry of group 14 element(ii) hydride complexes has rapidly expanded since the first stable example of such a compound was reported in 2000. Since that time it has become apparent that these systems display remarkable reactivity patterns, in some cases mimicking those of late transition-metal (TM) hydride compounds. This is especially so for the hydroelementation of unsaturated organic substrates. Recently, this aspect of their reactivity has been extended to the use of group 14 element(ii) hydrides as efficient, "TM-like" catalysts in organic synthesis. This review will detail how the chemistry of these hydride compounds has advanced since their early development. Throughout, there is a focus on the importance of ligand effects in these systems, and how ligand design can greatly modify a coordinated complex's electronic structure, reactivity, and catalytic efficiency.

Hydrogen absorption-desorption properties of U 2Ti

NASA Astrophysics Data System (ADS)

Takuya, Yamamoto; Satoru, Tanaka; Michio, Yamawaki

1990-02-01

Hydrogen absorption-desorption properties of U 2Ti intermetallic compound was examined over the temperature range of 298 to 973 K and at hydrogen pressures below 10 5 Pa. It absorbs hydrogen up to 7.6 atoms per F.U. (formula unit) by two step reactions and hence each desorption isotherm is separated into two plateau regions. In the first plateau, a newly-found ternary hydride is formed, where the hydrogen concentration, cH, reaches 2.4 H atoms/F.U. In the second plateau, UH 3 is formed and cH reaches 7.6 H atoms/F.U. The specimen is disintegrated into fine powder in the second plateau, while in the first plateau the ternary hydride which was identified to be UTi 2H x, ( x = 4.8 to 6.2) showed high durability against powdering. It is predicted that UTi 2 can be suitable material for tritium storage.

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

NASA Astrophysics Data System (ADS)

Karahan, Aydın; Buongiorno, Jacopo

2010-01-01

An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Metallic Fuels Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn E.; Papesch, Cynthia A.; Burkes, Douglas E.

This is not a typical External Report--It is a Handbook. No Abstract is involved. This includes both Parts 1 and 2. The Metallic Fuels Handbook summarizes currently available information about phases and phase diagrams, heat capacity, thermal expansion, and thermal conductivity of elements and alloys in the U-Pu-Zr-Np-Am-La-Ce-Pr-Nd system. Although many sections are reviews and updates of material in previous versions of the Handbook [1, 2], this revision is the first to include alloys with four or more elements. In addition to presenting information about materials properties, the handbook attempts to provide information about how well each property is knownmore » and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data.« less

Pre- and post-irradiation characterization and properties measurements of ZrC coated surrogate TRISO particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vasudevamurthy, Gokul; Katoh, Yutai; Hunn, John D

2010-09-01

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in TRISO fuel particles for high temperature gas-cooled reactor fuels. Six sets of ZrC coated surrogate microsphere samples, fabricated by the Japan Atomic Energy Agency using the fluidized bed chemical vapor deposition method, were irradiated in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. These developmental samples available for the irradiation experiment were in conditions of either as-fabricated coated particles or particles that had been heat-treated to simulate the fuel compacting process. Five sets of samples were composed of nominally stoichiometricmore » compositions, with the sixth being richer in carbon (C/Zr = 1.4). The samples were irradiated at 800 and 1250 C with fast neutron fluences of 2 and 6 dpa. Post-irradiation, the samples were retrieved from the irradiation capsules followed by microstructural examination performed at the Oak Ridge National Laboratory's Low Activation Materials Development and Analysis Laboratory. This work was supported by the US Department of Energy Office of Nuclear Energy's Advanced Gas Reactor program as part of International Nuclear Energy Research Initiative collaboration with Japan. This report includes progress from that INERI collaboration, as well as results of some follow-up examination of the irradiated specimens. Post-irradiation examination items included microstructural characterization, and nanoindentation hardness/modulus measurements. The examinations revealed grain size enhancement and softening as the primary effects of both heat-treatment and irradiation in stoichiometric ZrC with a non-layered, homogeneous grain structure, raising serious concerns on the mechanical suitability of these particular developmental coatings as a replacement for SiC in TRISO fuel. Samples with either free carbon or carbon-rich layers dispersed in the ZrC coatings experienced negligible grain size

Integrated Chemical Fuel Microprocessor for Power Generation in MEMS Applications

DTIC Science & Technology

2005-07-01

unreacted fuels (ammonia and hydrocarbon) and carbon monoxide that could otherwise adversely affect hydrogen Proton Exchange Membrane ( PEM ) fuel cell ...High hydrogen purity is required in a variety of processes, from the microelectronics industry to PEM fuel cells . For portable-power applications, it...Geff Ffuel Heat Load Complexity Li-Ion Batteries 330 140 1.2 W Low Carnot Engines *7,878 13,750 10% 50% 395 690 10 W Low Fuel Cells : PEM /Hydride #2,382

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less