Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2015-06-15

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validationmore » study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.« less

Powder formation of {gamma} uranium-molybdenum alloys via hydration-dehydration

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

Vaz de Oliveira, Fabio Branco; Durazzo, Michelangelo; Fontenele Urano de Carvalho, Elita

2008-07-15

Gamma uranium-molybdenum alloys has been considered as fuel phase in plate type fuel elements for MTR reactors, mainly due to their acceptable performance under irradiation and metallurgical processing. To its use as a dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and one of the techniques considered at IPEN / CNEN - Brazil is HDH (hydration-dehydration). The alloys were produced by the induction melting technique, and samples were obtained from the alloys for the thermal treatments, under constant flow of hydrogen, for temperatures varying from 400 deg C to 600more » deg C and times from 1 to 4 hours, followed by dehydration. A preliminary characterization of the powders was made and the curves of mass variation versus time were obtained and related to the powder characteristics. This paper describes the first results on the development of the technology to the powder formation of the (5 to 10) % weight molybdenum {gamma}-UMo alloys, and discusses some of its aspects, mainly those related to the {gamma} {yields} {alpha} equilibrium data. (author)« less

Application of the DART Code for the Assessment of Advanced Fuel Behavior

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

Rest, J.; Totev, T.

2007-07-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO{sub 2}more » fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)« less

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U-Mo/Al dispersion type fuel plates

NASA Astrophysics Data System (ADS)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

2014-10-01

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

Elevated Temperature Tensile Tests on DU–10Mo Rolled Foils

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

Schulthess, Jason

2014-09-01

Tensile mechanical properties for uranium-10 wt.% molybdenum (U–10Mo) foils are required to support modeling and qualification of new monolithic fuel plate designs. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low enriched U–10Mo to be used in the actual fuel plates, therefore DU-10Mo was studied to simplify material processing, handling, and testing requirements. In this report, tensile testing of DU-10Mo fuel foils prepared using four different thermomechanical processing treatments were conducted to assess the impact of foil fabrication history on resultant tensile properties.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Molybdenum-base cermet fuel development

NASA Astrophysics Data System (ADS)

Pilger, James P.; Gurwell, William E.; Moss, Ronald W.; White, George D.; Seifert, David A.

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. The MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermte. This cermet is to have a high matrix density (greater than or equal to 95 percent) for high strength and high thermal conductance coupled with a high particle (UN) porosity (approximately 25 percent) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous Un microspheres become available. Process development was conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and the vacuum hot press consolidation techniques. This paper summarizes the status of these activities.

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

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

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

DOEpatents

Clark, H.M.; Duffey, D.

1958-06-17

A process is described for extracting uranium from uranium ore, wherein the uranium is substantially free from molybdenum contamination. In a solvent extraction process for recovering uranium, uranium and molybdenum ions are extracted from the ore with ether under high acidity conditions. The ether phase is then stripped with water at a lower controiled acidity, resaturated with salting materials such as sodium nitrate, and reextracted with the separation of the molybdenum from the uranium without interference from other metals that have been previously extracted.

RECOVERY OF URANIUM FROM PITCHBLENDE

DOEpatents

Ruehle, A.E.

1958-06-24

The decontamination of uranium from molybdenum is described. When acid solutions containing uranyl nitrate are contacted with ether for the purpose of extracting the uranium values, complex molybdenum compounds are coextracted with the uranium and also again back-extracted from the ether with the uranium. This invention provides a process for extracting uranium in which coextraction of molybdenum is avoided. It has been found that polyhydric alcohols form complexes with molybdenum which are preferentially water-soluble are taken up by the ether extractant to only a very minor degree. The preferred embodiment of the process uses mannitol, sorbitol or a mixture of the two as the complexing agent.

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

NASA Astrophysics Data System (ADS)

Willingham, David; Naes, Benjamin E.; Tarolli, Jay G.; Schemer-Kohrn, Alan; Rhodes, Mark; Dahl, Michael; Guzman, Anthony; Burkes, Douglas E.

2018-01-01

Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

FUEL COMPOSITION FOR NUCLEAR REACTORS

DOEpatents

Andersen, J.C.

1963-08-01

A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

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

Rabin, S.A.; Lotts, A.L.; Hammond, J.P.

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodyingmore » hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)« less

Recovering and recycling uranium used for production of molybdenum-99

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

Reilly, Sean Douglas; May, Iain; Copping, Roy

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated targetmore » suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.« less

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

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

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.« less

Manufacturing Experience for Oxide Dispersion Strengthened Alloys

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

This report documents the results of the development and the manufacturing experience gained at the Pacific Northwest National Laboratories (PNNL) while working with the oxide dispersion strengthened (ODS) materials MA 956, 14YWT, and 9YWT. The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. ODS materials have the potential to provide improved performance for the U-Mo concept.

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

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

Reilly, Sean Douglas; May, Iain; Copping, Roy

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted tomore » concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.« less

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

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

B.R. Westphal; D. Vaden; S.X. Li

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontallymore » displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.« less

UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

NASA Astrophysics Data System (ADS)

Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

2018-02-01

The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

WHETSTONE ROADLESS AREA, ARIZONA.

USGS Publications Warehouse

Wrucke, Chester T.; McColly, Robert A.

1984-01-01

A mineral survey conducted has shown that areas in and adjacent to the Whetstone Roadless Area, Arizona have a substantiated resource potential for copper, lead, gold, silver, and quartz, and a probable mineral-resource potential for copper silver, lead, gold, molybdenum, tungsten, uranium, and gypsum. Copper and silver occur in a small vein deposit in the southwestern part of the roadless area. Copper, lead, silver, gold, and molybdenum are known in veins associated with a porphyry copper deposit in a reentrant near the southern border of the roadless area. Vein deposits of tungsten and uranium are possible in the northeast part of the roadless area near areas of known production of these commodities. Demonstrated resources of quartz for smelter flux extend into the roadless area from the Ricketts mine. Areas of probable potential for gypsum resources also occur within the roadless area. No potential for fossil fuel resources was identified in the study.

THE CHEMICAL ANALYSIS OF TERNARY ALLOYS OF PLUTONIUM WITH MOLYBDENUM AND URANIUM

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

Phillips, G.; Woodhead, J.; Jenkins, E.N.

1958-09-01

It is shown that the absorptiometric determination of molybdenum as thiocyanate may be used in the presence of plutonium. Molybdenum interferes with previously published methods for determining uranium and plutonium but conditlons have been established for its complete removal by solvent extraction of the compound with alpha -benzoin oxime. The previous methods for uranium and plutonium are satisfactory when applied to the residual aqueous phase following this solvent extraction. (auth)

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

NASA Technical Reports Server (NTRS)

Diianni, D. C.

1972-01-01

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

NPF MECHANICAL CELL NaK DISPOSAL AND FUME ABATEMENT

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

Rey, G.

Some of the fuels originally scheduled for processing in the nonproduction fuel (NPF) processing program incorporated sodium or sodium- potassium alloy (NaK) as the bonding material between stainless-steel cladding and the uranium or uranium-molybdenum alloy core. Because of the special hazards involved in handling NaK, studies were made to determine safe methods for processing NaK-containing fuels. An underwater NaK dispensing system was installed, and tests were made to determine the characteristics of the NaK-water reaction. The equipment consisted of a dispenser, reaction pan, and off-gas scrubber. After initinl studies, a prototype test was made wherein U-Mo canned slugs containing NaKmore » reservoirs were hack sawed underwater. The studies demonstrated that the NaK reservoirs can be safely deactivated by hack sawing under a submerged hood in a shallow water bath. (W.L.H.)« less

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

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

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

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.

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.

A physical description of fission product behavior fuels for advanced power reactors.

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

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

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

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop 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. The program is 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 layermore » 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.« 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

PROCESS OF MAKING SHAPED FUEL FOR NUCLEAR REACTORS

DOEpatents

O'Leary, W.J.; Fisher, E.A.

1964-02-11

A process for making uranium dioxide fuel of great strength, density, and thermal conductivity by mixing it with 0.1 to 1% of a densifier oxide (tin, aluminum, zirconium, ferric, zinc, chromium, molybdenum, titanium, or niobium oxide) and with a plasticizer (0.5 to 3% of bentonite and 0.05 to 2% of methylcellulose, propylene glycol alginate, or ammonium alginate), compacting the mixture obtained, and sintering the bodies in an atmosphere of carbon monoxide or carbon dioxide, with or without hydrogen, or of a nitrogen-hydrogen mixture is described. (AEC)

On the use of thermal NF3 as the fluorination and oxidation agent in treatment of used nuclear fuels

NASA Astrophysics Data System (ADS)

Scheele, Randall; McNamara, Bruce; Casella, Andrew M.; Kozelisky, Anne

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as a fluorination or fluorination/oxidation agent for separating valuable constituents from used nuclear fuels by exploiting the different volatilities of the constituent fission product and actinide fluorides. Our thermodynamic calculations show that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from oxides and metals that can form volatile fluorides. Simultaneous thermogravimetric and differential thermal analyses show that the oxides of lanthanum, cerium, rhodium, and plutonium are fluorinated but do not form volatile fluorides when treated with nitrogen trifluoride at temperatures up to 550 °C. However, depending on temperature, volatile fluorides or oxyfluorides can form from nitrogen trifluoride treatment of the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. Thermoanalytical studies demonstrate near-quantitative separation of uranium from plutonium in a mixed 80% uranium and 20% plutonium oxide. Our studies of neat oxides and metals suggest that the reactivity of nitrogen trifluoride may be adjusted by temperature to selectively separate the major volatile fuel constituent uranium from minor volatile constituents, such as Mo, Tc, Ru and from the non-volatile fuel constituents based on differences in their reaction temperatures and kinetic behaviors. This reactivity is novel with respect to that reported for other fluorinating reagents F2, BrF5, ClF3.

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

RERTR-13 Irradiation Summary Report

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

D. M. Perez; M. A. Lillo; G. S. Chang

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

PURIFICATION PROCESS

DOEpatents

Wibbles, H.L.; Miller, E.I.

1958-01-14

This patent deals with the separation of uranium from molybdenum compounds, and in particular with their separation from ether solutions containing the molybdenum in the form of acids, such as silicomolybdic and phosphomolybdic acids. After the nitric acid leach of pitchblende, the molybdenum values present in the ore are found in the leach solution in the form of complex acids. The uranium bearing solution may be purified of this molybdenum content by comtacting it with activated charcoal. The purification is improved when the acidity of the solution is low ad agitation is also beneficial. The molybdenum may subsequently be recovered from the charcosl ad the charcoal reused.

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

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

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

On the Use of Thermal NF3 as the Fluorination and Oxidation Agent in Treatment of Used Nuclear Fuels

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

Scheele, Randall D.; McNamara, Bruce K.; Casella, Andrew M.

2012-05-01

This paper presents results of our investigation on the use of nitrogen trifluoride as the fluorination or fluorination/oxidation agent for use in a process for separating valuable constituents from used nuclear fuels by employing the volatility of many transition metal and actinide fluorides. Nitrogen trifluoride is less chemically and reactively hazardous than the hazardous and aggressive fluorinating agents used to prepare uranium hexafluoride and considered for fluoride volatility based nuclear fuels reprocessing. In addition, nitrogen trifluoride’s less aggressive character may be used to separate the volatile fluorides from used fuel and from themselves based on the fluorination reaction’s temperature sensitivitymore » (thermal tunability) rather than relying on differences in sublimation/boiling temperature and sorbents. Our thermodynamic calculations found that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from candidate oxides and metals. Our simultaneous thermogravimetric and differential thermal analyses found that the oxides of lanthanum, cerium, rhodium, and plutonium fluorinated but did not form volatile fluorides and that depending on temperature volatile fluorides formed from the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. We also demonstrated near-quantitative removal of uranium from plutonium in a mixed oxide.« less

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

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

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 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. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

Vapor core propulsion reactors

NASA Technical Reports Server (NTRS)

Diaz, Nils J.

1991-01-01

Many research issues were addressed. For example, it became obvious that uranium tetrafluoride (UF4) is a most preferred fuel over uranium hexafluoride (UF6). UF4 has a very attractive vaporization point (1 atm at 1800 K). Materials compatible with UF4 were looked at, like tungsten, molybdenum, rhenium, carbon. It was found that in the molten state, UF4 and uranium attacked most everything, but in the vapor state they are not that bad. Compatible materials were identified for both the liquid and vapor states. A series of analyses were established to determine how the cavity should be designed. A series of experiments were performed to determine the properties of the fluid, including enhancement of the electrical conductivity of the system. CFD's and experimental programs are available that deal with most of the major issues.

MOUNT ZIRKEL WILDERNESS AND VICINITY, COLORADO.

USGS Publications Warehouse

Snyder, George L.; Patten, Lowell L.

1984-01-01

Several areas of metallic and nonmetallic mineralization have been identified from surface occurrences within the Mount Zirkel Wilderness and vicinity, Colorado. Three areas of probable copper-lead-zinc-silver-gold resource potential, two areas of probable chrome-platinum resource potential, four areas of probable uranium-thorium resource potential, two areas of probable molybdenum resource potential, and one area of probable fluorspar potential were identified. No potential for fossil fuel or geothermal resources was identified.

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

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

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

METHOD OF MAKING FUEL BODIES

DOEpatents

Goeddel, W.V.; Simnad, M.T.

1963-04-30

This patent relates to a method of making a fuel compact having a matrix of carbon or graphite which carries the carbides of fissile material. A nuclear fuel material selected from the group including uranium and thorium carbides, silicides, and oxides is first mixed both with sufficient finely divided carbon to constitute a matrix in the final product and with a diffusional bonding material selected from the class consisting of zirconium, niobium, molybdenum, titanium, nickel, chromium, and silicon. The mixture is then heated at a temperature of 1500 to 1800 nif- C while maintaining it under a pressure of over about 2,000 pounds per square inch. Preferably, heating is accomplished by the electrical resistance of the compact itself. (AEC)

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

High-Resolution Characterization of UMo Alloy Microstructure

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

Devaraj, Arun; Kovarik, Libor; Joshi, Vineet V.

2016-11-30

This report highlights the capabilities and procedure for high-resolution characterization of UMo fuels in PNNL. Uranium-molybdenum (UMo) fuel processing steps, from casting to forming final fuel, directly affect the microstructure of the fuel, which in turn dictates the in-reactor performance of the fuel under irradiation. In order to understand the influence of processing on UMo microstructure, microstructure characterization techniques are necessary. Higher-resolution characterization techniques like transmission electron microscopy (TEM) and atom probe tomography (APT) are needed to interrogate the details of the microstructure. The findings from TEM and APT are also directly beneficial for developing predictive multiscale modeling tools thatmore » can predict the microstructure as a function of process parameters. This report provides background on focused-ion-beam–based TEM and APT sample preparation, TEM and APT analysis procedures, and the unique information achievable through such advanced characterization capabilities for UMo fuels, from a fuel fabrication capability viewpoint.« less

Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

NASA Astrophysics Data System (ADS)

Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

2009-03-01

Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

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

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

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

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

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Processing of LEU targets for {sup 99}Mo production--testing and modification of the Cintichem process

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

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on {sup 99}Mo recovery and purification by its precipitation with {alpha}-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of {alpha}-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the {sup 99}Mo recoverymore » and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible.« less

Post-irradiation examination of uranium 7 wt% molybdenum atomized dispersion fuel

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Jarousse, C.; Huet, F.; Trotabas, M.; Boyard, M.; Guillot, S.; Sannen, L.; Verwerft, M.

2004-10-01

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm 2 while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al 3 and (U,Mo)Al 4. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l'énergie atomique (CEA).

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

Rabin, S.A.; Martin, M.M.; Lotts, A.L.

The fabricability of dispersion fuels using UO/sub 2/ or UC as the dispersoid and uranium combined with 10 to 15 wt% Mo as the matrix was investigated. Cores containing l7.8 wt% UO/sub 2/ dispersed in U-- 15 wt.% Mo were successfully fabricated to about 80% of theoretical density by cold pressing at 50 tsi, sintering at 1100 deg C, and cold coining at 50 tsi. Comparable results were obtained with UC as the dispersoid. Core fabrication results varied greatly with the type of matrix powder used. Occluded gases, pour density, and surface cleanliness bore important relations to the fabrication behaviormore » of powders. Suitable pressing and sintering results were obtained with prealloyed, calcium-reduced U--Mo powder and with molybdenum and calcium-reduced uranium as elemental powders. Shotted prealloyed powders were difficult to press and sinter, as were elemental and prealloyed powders prepared by hydriding. The cores containing UO/sub 2/ were picture-frame, hot-roll-clad as miniature plates. Molybdenum, Fansteel 82, and Zr--3 wt% Al were investigated as cladding materials. While each bonded well to itself, only the molybdenum-clad core, rolled at 1150 deg C to 10/1 reduction, resulted in dispersions free of ruptures and UO/sub 2/ fragmentation and in strong bonding to the core, evaluated by metallography, mechanical peel, and thermal shock tests. The matrix phase was homogeneous, but the UO/sub 2/ dispersoid showed stringering characteristic of cores worked by hot rolling. Core densities as high as 99% of theoretical were obtained. (auth)« less

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors.

PubMed

Westman, Bjorn; Miller, Brandon; Jue, Jan-Fong; Aitkaliyeva, Assel; Keiser, Dennis; Madden, James; Tucker, Julie D

2018-07-01

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements. Copyright © 2018 Elsevier Ltd. All rights reserved.

RERTR-12 Insertion 2 Irradiation Summary Report

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

D. M. Perez; G. S. Chang; D. M. Wachs

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

THE GRAVIMETRIC DETERMINATION OF MOLYBDENUM IN URANIUM-MOLYBDENUM ALLOYS

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

None

1959-03-01

The sample is dissolved in nitric and hydrochloric acids. After heating the solution with sulfuric acid, molybodenum is precipitated as the benzoin-oxime complex which is ignited to molybdic oxide. This is dissolved in ammonia, and the molybdenum is precipitated and weighed as lead molybdate. (auth)

Predicting thermo-mechanical behaviour of high minor actinide content composite oxide fuel in a dedicated transmutation facility

NASA Astrophysics Data System (ADS)

Lemehov, S. E.; Sobolev, V. P.; Verwerft, M.

2011-09-01

The European Facility for Industrial Transmutation (EFIT) of the minor actinides (MA), from LWR spent fuel is being developed in the integrated project EUROTRANS within the 6th Framework Program of EURATOM. Two composite uranium-free fuel systems, containing a large fraction of MA, are proposed as the main candidates: a CERCER with magnesia matrix hosting (Pu,MA)O 2-x particles, and a CERMET with metallic molybdenum matrix. The long-term thermal and mechanical behaviour of the fuel under the expected EFIT operating conditions is one of the critical issues in the core design. To make a reliable prediction of long-term thermo-mechanical behaviour of the hottest fuel rods in the lead-cooled version of EFIT with thermal power of 400 MW, different fuel performance codes have been used. This study describes the main results of modelling the thermo-mechanical behaviour of the hottest CERCER fuel rods with the fuel performance code MACROS which indicate that the CERCER fuel residence time can safely reach at least 4-5 effective full power years.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

Hydrogeochemical and stream sediment detailed geochemical survey for Edgemont, South Dakota; Wyoming

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

Butz, T.R.; Dean, N.E.; Bard, C.S.

1980-05-31

Results of the Edgemont detailed geochemical survey are reported. Field and laboratory data are presented for 109 groundwater and 419 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are given. A generalized geologic map of the survey area is provided, and pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. Groundwaters containing greater than or equal to 7.35 ppB uranium are present in scattered clusters throughout the area sampled. Most of these groundwaters are from wells drilled where the Inyan Kara Group is exposed at themore » surface. The exceptions are a group of samples in the northwestern part of the area sampled and south of the Dewey Terrace. These groundwaters are also produced from the Inyan Kara Group where it is overlain by the Graneros Group and alluvium. The high uranium groundwaters along and to the south of the terrace are characterized by high molybdenum, uranium/specific conductance, and uranium/sulfate values. Many of the groundwaters sampled along the outcrop of the Inyan Kara Group are near uranium mines. Groundwaters have high amounts of uranium and molybdenum. Samples taken downdip are sulfide waters with low values of uranium and high values of arsenic, molybdenum, selenium, and vanadium. Stream sediments containing greater than or equal to 5.50 ppM soluble uranium are concentrated in basins draining the Graneros and Inyan Kara Groups. These values are associated with high values for arsenic, selenium, and vanadium in samples from both groups. Anomalous values for these elements in the Graneros Group may be caused by bentonite beds contained in the rock units. As shown on the geochemical distribution plot, high uranium values that are located in the Inyan Kara Group are almost exclusively draining open-pit uranium mines.« less

DEFLECTION OF A HETEROGENEOUS WIDE-BEAM UNDER UNIFORM PRESSURE LOAD

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

T. V. Holschuh; T. K. Howard; W. R. Marcum

2014-07-01

Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or generic test plate assembly (GTPA), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates onset by hydraulic forces. This test program supports ongoing work conducted for/by the Global Threat Reduction Initiative (GTRI) Fuels Development Program. This study’s focus supports the ongoing collaborative effort by detailing the derivation of an analytic solution for deflection of a heterogeneousmore » plate under a uniform, distributed load in order to predict the deflection of test plates in the GTPA. The resulting analytical solutions for three specific boundary condition sets are then presented against several test cases of a homogeneous plate. In all test cases considered, the results for both homogeneous and heterogeneous plates are numerically identical to one another, demonstrating correct derivation of the heterogeneous solution. Two additional problems are presents herein that provide a representative deflection profile for the plates under consideration within the GTPA. Furthermore, qualitative observations are made about the influence of a more-rigid internal fuel-meat region and its influence on the overall deflection profile of a plate. Present work is being directed to experimentally confirm the analytical solution’s results using select materials.« less

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

DOE PAGES

Hames, Amber L.; Tkac, Peter; Paulenova, Alena; ...

2017-01-17

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate themore » feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« less

Hot rolling of thick uranium molybdenum alloys

DOEpatents

DeMint, Amy L.; Gooch, Jack G.

2015-11-17

Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

TEM analysis of irradiation-induced interaction layers in coated UMo/X/Al trilayer systems (X= Ti, Nb, Zr, and Mo)

NASA Astrophysics Data System (ADS)

Chiang, H.-Y.; Wiss, T.; Park, S.-H.; Dieste-Blanco, O.; Petry, W.

2018-02-01

Uranium-molybdenum (UMo) alloy powder embedded in an Al matrix is considered as a promising candidate for fuel conversion of research reactors. A modified system with a diffusion barrier X as coating, UMo/X/Al trilayer (X = Ti, Zr, Nb, and Mo), has been investigated to suppress interdiffusion between UMo and the Al matrix. The trilayer systems were tested by swift heavy ion irradiation, the thereby created interaction zone has been analyzed by scanning transmission electron microscopy (STEM) and energy-dispersive X-ray spectroscopy (EDX). Detailed structural characterization are presented and compared to earlier μ-XRD analysis.

Effect of molybdenum ion implantation of the pitting corrosion of depleted uranium - 0.75 titanium alloy. (Reannouncement with new availability information). Final report

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

Lei, K.S.; Chang, F.; Levy, M.

1993-07-01

Pitting corrosion of molybdenum-ion-implanted, depleted uranium -0 75 Ti (DU -0 75 Ti) has been studied electrochemically in acidic, neutral, and alkaline solutions containing sodium chloride, and the results have been compared to those of the unimplanted DU -0 75 Ti. The data show that Mo implantation shifts the pitting potential of DU -0 75 Ti in the noble direction in acidic and alkaline solutions. In neutral 50 ppm Cl- solution, however, there is no beneficial effect of Mo implantation. Auger analysis studies show that before exposure to the solutions, all the molybdenum is in the oxide, which is approximatelymore » l000 A thick. After electrochemical scans in the acidic and alkaline chloride solutions, most of the Mo disappears from the oxide. However, no decrease in Mo concentration is found after exposure in neutral chloride solution. It is proposed that the implanted molybdenum dissolves in the acidic and alkaline solutions and forms simple or complex molybdates that inhibit pitting corrosion. The implanted molybdenum does not dissolve in the neutral chloride solution and inhibition does not occur.« less

Procedure for Uranium-Molybdenum Density Measurements and Porosity Determination

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

Prabhakaran, Ramprashad; Devaraj, Arun; Joshi, Vineet V.

2016-08-13

The purpose of this document is to provide guidelines for preparing uranium-molybdenum (U-Mo) specimens, performing density measurements, and computing sample porosity. Typical specimens (solids) will be sheared to small rectangular foils, disks, or pieces of metal. A mass balance, solid density determination kit, and a liquid of known density will be used to determine the density of U-Mo specimens using the Archimedes principle. A standard test weight of known density would be used to verify proper operation of the system. By measuring the density of a U-Mo sample, it is possible to determine its porosity.

COATING URANIUM FROM CARBONYLS

DOEpatents

Gurinsky, D.H.; Storrs, S.S.

1959-07-14

Methods are described for making adherent corrosion resistant coatings on uranium metal. According to the invention, the uranium metal is heated in the presence of an organometallic compound such as the carbonyls of nickel, molybdenum, chromium, niobium, and tungsten at a temperature sufficient to decompose the metal carbonyl and dry plate the resultant free metal on the surface of the uranium metal body. The metal coated body is then further heated at a higher temperature to thermally diffuse the coating metal within the uranium bcdy.

Uptake of trace elements and radionuclides from uranium mill tailings by four-wing saltbush (Atriplex canescens) and alkali sacaton (Sporobolus airoides). [Radium 226; Uranium; Molybdenum; Selenium; Vanadium; Astatine

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

Dreesen, D.R.; Marple, M.L.

1979-01-01

A greenhouse experiment was performed to determine the uptake of trace elements and radionuclides from uranium mill tailings by native plant species. Four-wing saltbush and alkali sacaton were grown in alkaline tailings covered with soil and in soil alone as controls. The tailings material was highly enriched in Ra-226, Mo, U, Se, V, and As compared with three local soils. The shrub grown in tailings had elevated concentrations of Mo, Se, Ra-226, U, As, and Na compared with the controls. Alkali sacaton contained high concentrations of Mo, Se, Ra-226, and Ni when grown on tailings. Molybdenum and selenium concentrations inmore » plants grown in tailings are above levels reported to be toxic to grazing animals. These results indicate that the bioavailability of Mo and Se in alkaline environments makes these elements among the most hazardous contaminants present in uranium mill wastes.« less

URANIUM ALLOYS

DOEpatents

Seybolt, A.U.

1958-04-15

Uranium alloys containing from 0.1 to 10% by weight, but preferably at least 5%, of either zirconium, niobium, or molybdenum exhibit highly desirable nuclear and structural properties which may be improved by heating the alloy to about 900 d C for an extended period of time and then rapidly quenching it.

VASQUEZ PEAK WILDERNESS STUDY AREA, AND ST. LOUIS PEAK, AND WILLIAMS FORK ROADLESS AREAS, COLORADO.

USGS Publications Warehouse

Theobald, P.K.; Bielski, A.M.

1984-01-01

A mineral-resource survey was conducted during the years 1979-82 in the Vasquez Peak Wilderness Study Area and in the St. Louis Peak and Williams Fork Roadless Areas, central Front Range, Colorado. Probable resource potential for the occurrence of copper, lead, zinc, and silver in massive sulfide deposits has been identified in calcareous metamorphic rocks in the northern part of the St. Louis Peak Roadless Area and in the southern part of the Williams Fork Roadless Area. A probable resource potential for vein-type uranium deposits is identified along the Berthoud Pass fault zone in the eastern part of the Vasquez Peak Wilderness Study Area. A large area encompassing the eastern and southeastern part of each of the three areas has probable and substantiated potential for either high-grade lead-zinc-silver vein deposits, or larger, lower-grade clustered vein deposits. A probable resource potential for stockwork molybdenum deposits related to porphyry molybdenum type mineralization exists beneath the lead-zinc-silver-rich veins. The nature of the geologic terrane indicates little likelihood for the occurrence of organic fuels.

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

Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

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

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing highmore » enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.« less

Enhancing the corrosion resistance of the 2205 duplex stainless steel bipolar plates in PEMFCs environment by surface enriched molybdenum

NASA Astrophysics Data System (ADS)

Jinlong, Lv; Zhuqing, Wang; Tongxiang, Liang; Ken, Suzuki; Hideo, Miura

Surface molybdenum enrichment on 2205 duplex stainless steel was obtained by the ball milling technique. The electrochemical results showed molybdenum enrichment on the surface of 2205 duplex stainless steel improved its corrosion resistance in a typical proton exchange membrane fuel cell environment. This was mainly attributed to higher molybdenum content in the passive film formed on 2205 duplex stainless steel after ball milling. The decreased donor and acceptor concentrations improved significantly the corrosion resistance of surface molybdenum-enriched 2205 duplex stainless steel bipolar plates in the simulated cathodic proton exchange membrane fuel cells environment. In addition, the interfacial contact resistance of the 2205 duplex stainless steel bipolar plates slightly decreased due to surface molybdenum enrichment.

Hydrothermal uranium deposits containing molybdenum and fluorite in the Marysvale volcanic field, west-central Utah

USGS Publications Warehouse

Cunningham, C.G.; Rasmussen, J.D.; Steven, T.A.; Rye, R.O.; Rowley, P.D.; Romberger, S.B.; Selverstone, J.

1998-01-01

Uranium deposits containing molybdenum and fluorite occur in the Central Mining Area, near Marysvale, Utah, and formed in an epithermal vein system that is part of a volcanic/hypabyssal complex. They represent a known, but uncommon, type of deposit; relative to other commonly described volcanic-related uranium deposits, they are young, well-exposed and well-documented. Hydrothermal uranium-bearing quartz and fluorite veins are exposed over a 300 m vertical range in the mines. Molybdenum, as jordisite (amorphous MoS2, together with fluorite and pyrite, increase with depth, and uranium decreases with depth. The veins cut 23-Ma quartz monzonite, 20-Ma granite, and 19-Ma rhyolite ash-flow tuff. The veins formed at 19-18 Ma in a 1 km2 area, above a cupola of a composite, recurrent, magma chamber at least 24 ?? 5 km across that fed a sequence of 21- to 14-Ma hypabyssal granitic stocks, rhyolite lava flows, ash-flow tuffs, and volcanic domes. Formation of the Central Mining Area began when the intrusion of a rhyolite stock, and related molybdenite-bearing, uranium-rich, glassy rhyolite dikes, lifted the fractured roof above the stock. A breccia pipe formed and relieved magmatic pressures, and as blocks of the fractured roof began to settle back in place, flat-lying, concave-downward, 'pull-apart' fractures were formed. Uranium-bearing, quartz and fluorite veins were deposited by a shallow hydrothermal system in the disarticulated carapace. The veins, which filled open spaces along the high-angle fault zones and flat-lying fractures, were deposited within 115 m of the ground surface above the concealed rhyolite stock. Hydrothermal fluids with temperatures near 200??C, ??18OH2O ~ -1.5, ?? -1.5, ??DH2O ~ -130, log fO2 about -47 to -50, and pH about 6 to 7, permeated the fractured rocks; these fluids were rich in fluorine, molybdenum, potassium, and hydrogen sulfide, and contained uranium as fluoride complexes. The hydrothermal fluids reacted with the wallrock resulting in precipitation of uranium minerals. At the deepest exposed levels, wall-rocks were altered to sericite; and uraninite, coffinite, jordisite, fluorite, molybdenite, quartz, and pyrite were deposited in the veins. The fluids were progressively oxidized and cooled at higher levels in the system by boiling and degassing; iron-bearing minerals in wall rocks were oxidized to hematite, and quartz, fluorite, minor siderite, and uraninite were deposited in the veins. Near the ground surface, the fluids were acidified by condensation of volatiles and oxidation of hydrogen sulfide in near-surface, steam-heated, ground waters; wall rocks were altered to kaolinite, and quartz fluorite, and uraninite were deposited in veins. Secondary uranium minerals, hematite, and gypsum formed during supergene alteration later in the Cenozoic when the upper part of the mineralized system was exposed by erosion.

Interpretation and modelling of fission product Ba and Mo releases from fuel

NASA Astrophysics Data System (ADS)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

Characterization of Bond Strength of U-Mo Fuel Plates Using the Laser Shockwave Technique: Capabilities and Preliminary Results

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

J. A. Smith; D. L. Cottle; B. H. Rabin

2013-09-01

This report summarizes work conducted to-date on the implementation of new laser-based capabilities for characterization of bond strength in nuclear fuel plates, and presents preliminary results obtained from fresh fuel studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Characterization involves application of two complementary experimental methods, laser-shock testing and laser-ultrasonic imaging, collectively referred to as the Laser Shockwave Technique (LST), that allows the integrity, physical properties and interfacial bond strength in fuel plates to be evaluated. Example characterization results are provided, including measurement of layer thicknesses, elastic properties ofmore » the constituents, and the location and nature of generated debonds (including kissing bonds). LST provides spatially localized, non-contacting measurements with minimum specimen preparation, and is ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterizing nuclear fuel plates are described, and preliminary bond strength measurement results are discussed, with emphasis on demonstrating the capabilities and limitations of these methods. These preliminary results demonstrate the ability to distinguish bond strength variations between different fuel plates. Although additional development work is necessary to validate and qualify the test methods, these results suggest LST is viable as a method to meet fuel qualification requirements to demonstrate acceptable bonding integrity.« less

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

NASA Astrophysics Data System (ADS)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Geochemical reconnaissance study of Vassar Meadow (Adams Rib) wetlands and vicinity, Eagle County, Colorado

USGS Publications Warehouse

Owen, Douglass E.; Breit, George N.

1995-01-01

Wetlands are known to be efficient filters of metals dissolved in ground and surface waters. This paper presents the results of geochemical reconnaissance sampling done at the request of the U.S. Environmental Protection Agency in wetlands in Vassar Meadow, Eagle County, Colorado. Ten wetlands were sampled and found to be variously enriched in chromium, molybdenum, and uranium. The uranium and chromium concentrations (and, to a lesser extent, molybdenum) represent an environmental concern should they be released as a result of anthropogenic disturbance. The metal accumulation in these wetlands documents that the wetlands have been functioning as filters that protect water quality in East Brush Creek by lowering the dissolved metal content in water.

Method of fabricating a uranium-bearing foil

DOEpatents

Gooch, Jackie G [Seymour, TN; DeMint, Amy L [Kingston, TN

2012-04-24

Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.

Metals in Urine and Diabetes in U.S. Adults

PubMed Central

Guallar, Eliseo; Cowie, Catherine C.

2016-01-01

Our objective was to evaluate the relationship of urine metals including barium, cadmium, cobalt, cesium, molybdenum, lead, antimony, thallium, tungsten, and uranium with diabetes prevalence. Data were from a cross-sectional study of 9,447 participants of the 1999–2010 National Health and Nutrition Examination Survey, a representative sample of the U.S. civilian noninstitutionalized population. Metals were measured in a spot urine sample, and diabetes status was determined based on a previous diagnosis or an A1C ≥6.5% (48 mmol/mol). After multivariable adjustment, the odds ratios of diabetes associated with the highest quartile of metal, compared with the lowest quartile, were 0.86 (95% CI 0.66–1.12) for barium (Ptrend = 0.13), 0.74 (0.51–1.09) for cadmium (Ptrend = 0.35), 1.21 (0.85–1.72) for cobalt (Ptrend = 0.59), 1.31 (0.90–1.91) for cesium (Ptrend = 0.29), 1.76 (1.24–2.50) for molybdenum (Ptrend = 0.01), 0.79 (0.56–1.13) for lead (Ptrend = 0.10), 1.72 (1.27–2.33) for antimony (Ptrend < 0.01), 0.76 (0.51–1.13) for thallium (Ptrend = 0.13), 2.18 (1.51–3.15) for tungsten (Ptrend < 0.01), and 1.46 (1.09–1.96) for uranium (Ptrend = 0.02). Higher quartiles of barium, molybdenum, and antimony were associated with greater HOMA of insulin resistance after adjustment. Molybdenum, antimony, tungsten, and uranium were positively associated with diabetes, even at the relatively low levels seen in the U.S. population. Prospective studies should further evaluate metals as risk factors for diabetes. PMID:26542316

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, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.« less

Distribution of selenium, molybdenum and uranium in sediment cores from the Colorado River delta, Baja California, Mexico.

PubMed

Orozco-Durán, A; Daesslé, L W; Gutiérrez-Galindo, E A; Muñoz-Barbosa, A

2012-01-01

The distribution of selenium, molybdenum and uranium was studied in ~1.5 m sediment cores from the Colorado River delta, at the Colorado (CR) and Hardy (HR) riverbeds. Core HR2 showed highest Se, Mo and U concentrations at its bottom (2.3, 0.95 and 1.8 μg g(-1)) within a sandy-silt layer deposited prior to dam construction. In CR5 the highest concentrations of these elements (0.9, 1.4 and 1.7 μg g(-1) respectively) were located at the top of the core within a surface layer enriched in organic carbon. A few samples from HR2 had Se above the probable toxic effect level guidelines.

Surface engineering of low enriched uranium-molybdenum

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Detavernier, C.

2013-09-01

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

Distribution of trace elements in drilling chip samples around a roll-type uranium deposit, San Juan Basin, New Mexico

USGS Publications Warehouse

Day, H.C.; Spirakis, C.S.; Zech, R.S.; Kirk, A.R.

1983-01-01

Chip samples from rotary drilling in the vicinity of a roll-type uranium deposit in the southwestern San Juan Basin were split into a whole-washed fraction, a clay fraction, and a heavy mineral concentrate fraction. Analyses of these fractions determined that cutting samples could be used to identify geochemical halos associated with this ore deposit. In addition to showing a distribution of selenium, uranium, vanadium, and molybdenum similar to that described by Harshman (1974) in uranium roll-type deposits in Wyoming, South Dakota, and Texas, the chemical data indicate a previously unrecognized zinc anomaly in the clay fraction downdip of the uranium ore.

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

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

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

Molybdenum-UO2 cerment irradiation at 1145 K

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

76 FR 67765 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-02

... Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding Louisiana Energy Services, National..., Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety... Commission. Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards...

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

Fission gas bubble identification using MATLAB's image processing toolbox

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

Collette, R.; King, J.; Keiser, Jr., D.

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

Fission gas bubble identification using MATLAB's image processing toolbox

DOE PAGES

Collette, R.; King, J.; Keiser, Jr., D.; ...

2016-06-08

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

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 metal organic chemical precursors tetrakis dimethylamido zirconium (TDMAZr) and ammonia( NH3) for succesful deposition of ZrN coating. Nitrogen (N2) gas carried the chemicals to a hot wall reactor maintained at a temperature range of 235 to 245 °C. The ALD system design evolved over the course of this research as the process variables were steadily improved. The conditions found deemed for attaining best coating were at a temperature of 245 °C, with pulse time of 0.8 seconds for TDMAZr and 0.1 seconds for NH3 along with 15 seconds of purge time in-between each cycle. The ALD system was successful in making 1-micrometer (um) ZrN with low levels of chemical impurities over U-Mo powder batches. The deposited coatings were characterized using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), electron energy loss spectroscopy (EELS) and Transmission electron microscope (TEM). This document describes the establishment of the Savannah 200 ALD system, precursor surface reaction procedures and finally the nature of the coating achieved, including characterization of the coating at the different stages of deposition. It was found that an interlayer of alumina in between ZrN and the U-Mo surface was required to reduce the residual stress generated during the ALD procedure. The alumina not only removed the risk of cracking and spallation of the ZrN coating but also provided adequate strength for the barrier layer to withstand the fuel plate rolling conditions. The ZrN coating was nano crystalline in nature, with grain size varying from 5-10 nm, the deposited layer was found to be dense consisting of a layered structure. The coating could retain its crystallinity and maintain its phase when irradiated with 1 MeV single charged ion Kr to produce a damage of 10 displacement per atom (DPA) at intermediate voltage electron microscopy (IVEM).

Comparison of the Environment, Health, And Safety Characteristics of Advanced Thorium- Uranium and Uranium-Plutonium Fuel Cycles

NASA Astrophysics Data System (ADS)

Ault, Timothy M.

The environment, health, and safety properties of thorium-uranium-based (''thorium'') fuel cycles are estimated and compared to those of analogous uranium-plutonium-based (''uranium'') fuel cycle options. A structured assessment methodology for assessing and comparing fuel cycle is refined and applied to several reference fuel cycle options. Resource recovery as a measure of environmental sustainability for thorium is explored in depth in terms of resource availability, chemical processing requirements, and radiological impacts. A review of available experience and recent practices indicates that near-term thorium recovery will occur as a by-product of mining for other commodities, particularly titanium. The characterization of actively-mined global titanium, uranium, rare earth element, and iron deposits reveals that by-product thorium recovery would be sufficient to satisfy even the most intensive nuclear demand for thorium at least six times over. Chemical flowsheet analysis indicates that the consumption of strong acids and bases associated with thorium resource recovery is 3-4 times larger than for uranium recovery, with the comparison of other chemical types being less distinct. Radiologically, thorium recovery imparts about one order of magnitude larger of a collective occupational dose than uranium recovery. Moving to the entire fuel cycle, four fuel cycle options are compared: a limited-recycle (''modified-open'') uranium fuel cycle, a modified-open thorium fuel cycle, a full-recycle (''closed'') uranium fuel cycle, and a closed thorium fuel cycle. A combination of existing data and calculations using SCALE are used to develop material balances for the four fuel cycle options. The fuel cycle options are compared on the bases of resource sustainability, waste management (both low- and high-level waste, including used nuclear fuel), and occupational radiological impacts. At steady-state, occupational doses somewhat favor the closed thorium option while low-level waste volumes slightly favor the closed uranium option, although uncertainties are significant in both cases. The high-level waste properties (radioactivity, decay heat, and ingestion radiotoxicity) all significantly favor the closed fuel cycle options (especially the closed thorium option), but an alternative measure of key fission product inventories that drive risk in a repository slightly favors the uranium fuel cycles due to lower production of iodine-129. Resource requirements are much lower for the closed fuel cycle options and are relatively similar between thorium and uranium. In additional to the steady-state results, a variety of potential transition pathways are considered for both uranium and thorium fuel cycle end-states. For dose, low-level waste, and fission products contributing to repository risk, the differences among transition impacts largely reflected the steady-state differences. However, the HLW properties arrived at a distinctly opposite result in transition (strongly favoring uranium, whereas thorium was strongly favored at steady-state), because used present-day fuel is disposed without being recycled given that uranium-233, rather than plutonium, is the primarily fissile nuclide at the closed thorium fuel cycle's steady-state. Resource consumption was the only metric was strongly influenced by the specific transition pathway selected, favoring those pathways that more quickly arrived at steady-state through higher breeding ratio assumptions regardless of whether thorium or uranium was used.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

NASA Astrophysics Data System (ADS)

Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

2013-02-01

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

Synthesis of pentafluorides

DOEpatents

Asprey, L.B.; Paine, R.T. Jr.

1975-12-30

The reactions of uranium, molybdenum, rhenium, osmium and iridium hexafluorides with hydrogen gas in the presence of ultraviolet radiation or with silicon powder in an anhydrous HF slurry provide especially useful, high yield syntheses of pure pentafluorides.

Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

2011-12-01

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

NASA Astrophysics Data System (ADS)

Ignatiev, Victor; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

2013-09-01

In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF-BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80М-VI (Mo—12, Cr—7.6, Nb—1.5), HN80МТY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80МТW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and ЕМ-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility. If the redox state the fuel salt is characterized by uranium ratio [U(IV)]/[U(III)] < 1 the alloys' specimens get a more negative stationary electrode potential than equilibrium electrode potentials of some uranium intermetallic compounds and alloys with nickel and molybdenum. This leads to spontaneous behavior of alloy formation processes on the specimens' surface and further diffusion of uranium deep into the metallic phase. As consequence of this films of intermetallic compounds and alloys of nickel, molybdenum, tungsten with uranium are formed on the alloys specimens' surface, and intergranular corrosion does not take place. In the fuel salt with [U(IV)]/[U(III)] = 4-20 the potentials of uranium alloy formation with the main components of the tested alloys are not reached, that's why alloys and intermetallic compounds are not formed on the surface of the investigated chromium-nickel alloys. Under such conditions any intergranular tellurium corrosion of the selected alloys does not occur. In the fuel salt with [U(IV)/]/[U(III)] = 100 the potentials of uranium alloy formation with the main components of the tested alloys are not also reached. Under such redox conditions any traces intergranular tellurium IGC on the HN80MTY and H80M-VI alloys specimens are not found. Certain signs of incipient IGC in the form of tellurium presence on the grain boundaries in the HN80MTB and EM-721 alloys surface layer and formation of not too deep cracks on HN80MTB alloy surface were revealed at [U(IV)/]/[U(III)] = 100. With this uranium ratio in the presence of corrosion products on the surface of all of the alloys films, containing tellurium, metals of the construction alloys and carbon, are formed. In the melt with [U(IV)]/[U(III)] = 500 in all of the alloys tested the tellurium IGC took place. The HN80MTY alloy shows the maximum resistance to tellurium IGC. The intensity of tellurium IGC of the alloy (the K parameter) is by 3-5 times lower as compared to other alloys. The EM-721 alloy has the minimal resistance to tellurium IGC (K = 9200 pc m/cm, the depth of cracks is up to 434 μm). The studies have shown, that the intensity of the nickel alloys IGC is controlled by the [U(IV)]/[U(III)] ratio, and its dependence on this parameter is of threshold character. Providing the uranium ratio value's monitoring and regulation, it is possible to control the tellurium corrosion and in such a way to eliminate IGC completely or to minimize its value. The alloys strength characteristics and their structure were changed insignificantly after testing within the [U(IV)]/[U(III)] range from 0.7 tо 100. The changes are not linked with the influence of fuel salt, containing tellurium additions, but are stipulated by alloys structure, temperature factor, exposure time and mechanical loads. Significant effect of tellurium cracking on the alloys (excepting HN80MTY) strength characteristics was established after corrosion testing with [U(IV)]/[U(III)] = 500. In the absence of IGC all of the alloys investigated have a good ductility at high strength characteristics. The disrupture of specimens under mechanical tests both before and after corrosion tests of all alloys except for ЕМ-721 proceeds on a ductile mechanism. On the EM-721 alloy specimens, both in their initial state and after corrosion testing, clear signs of brittle destruction, caused by heterogeneity of its structure due to the presence of tungsten phase, are very clearly observed. The presence of such phases increases the alloy IGC and leads to reduction of the alloy resistance tellurium damage. The HN80MTY alloy has the best corrosion and mechanical properties. It does not undergo tellurium IGC in the molten 75LiF-5BeF2-20ThF4 salt mixture fueled by about 2 mol% of UF4 with [U(IV)]/[U(III)] ratio ⩽ 100. The alloy has high resistance to tellurium cracking at [U(IV)]/[U(III)] = 500. The alloy can be recommended as the main construction material for the fuel circuit with selected salt composition up to temperature 750 °С.

Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

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

B.R. Westphal; G.L. Fredrickson; G.G. Galbreth

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.« less

Pyroprocessing of fast flux test facility nuclear fuel

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

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)« less

Benefits of utilizing CellProfiler as a characterization tool for U-10Mo nuclear fuel

DOE PAGES

Collette, R.; Douglas, J.; Patterson, L.; ...

2015-05-01

Automated image processing techniques have the potential to aid in the performance evaluation of nuclear fuels by eliminating judgment calls that may vary from person-to-person or sample-to-sample. Analysis of in-core fuel performance is required for design and safety evaluations related to almost every aspect of the nuclear fuel cycle. This study presents a methodology for assessing the quality of uranium-molybdenum fuel images and describes image analysis routines designed for the characterization of several important microstructural properties. The analyses are performed in CellProfiler, an open-source program designed to enable biologists without training in computer vision or programming to automatically extract cellularmore » measurements from large image sets. The quality metric scores an image based on three parameters: the illumination gradient across the image, the overall focus of the image, and the fraction of the image that contains scratches. The metric presents the user with the ability to ‘pass’ or ‘fail’ an image based on a reproducible quality score. Passable images may then be characterized through a separate CellProfiler pipeline, which enlists a variety of common image analysis techniques. The results demonstrate the ability to reliably pass or fail images based on the illumination, focus, and scratch fraction of the image, followed by automatic extraction of morphological data with respect to fission gas voids, interaction layers, and grain boundaries.« less

Dosimetry characterization of the Godiva Reactor under burst conditions

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

Hickman, D. P.; Heinrichs, D. P.; Hudson, R.

2017-06-22

A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C,more » and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.« less

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

NASA Astrophysics Data System (ADS)

Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

High loading uranium fuel plate

DOEpatents

Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry R.

1990-01-01

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pair of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat having a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process.

REDUCTION OF INORGANIC COMPOUNDS WITH MOLECULAR HYDROGEN BY MICROCOCCUS LACTILYTICUS I.

PubMed Central

Woolfolk, C. A.; Whiteley, H. R.

1962-01-01

Woolfolk, C. A. (University of Washington, Seattle) and H. R. Whiteley. Reduction of inorganic compounds with molecular hydrogen by Micrococcus lactilyticus. I. Stoichiometry with compounds of arsenic, selenium, tellurium, transition and other elements. J. Bacteriol. 84:647–658. 1962.—Extracts of Micrococcus lactilyticus (Veillonella alcalescens) oxidize molecular hydrogen at the expense of certain compounds of arsenic, bismuth, selenium, tellurium, lead, thallium, vanadium, manganese, iron, copper, molybdenum, tungsten, osmium, ruthenium, gold, silver, and uranium, as well as molecular oxygen. Chemical and manometric data indicate that the following reductions are essentially quantitative: arsenate to arsenite, pentavalent and trivalent bismuth to the free element, selenite via elemental selenium to selenide, tellurate and tellurite to tellurium, lead dioxide and manganese dioxide to the divalent state, ferric to ferrous iron, osmium tetroxide to osmate ion, osmium dioxide and trivalent osmium to the metal, uranyl uranium to the tetravalent state, vanadate to the level of vanadyl, and polymolybdate ions to molybdenum blues with an average valence for molybdenum of +5. The results of a study of certain other hydrogenase-containing bacteria with respect to their ability to carry out some of the same reactions are also presented. PMID:14001842

Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

DOEpatents

Herrmann, Steven Douglas

2014-05-27

Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

DOEpatents

Buyers, A.G.

1959-06-30

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2012 CFR

2012-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2014 CFR

2014-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2013 CFR

2013-07-01

... SOURCE CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of... off-pounds) of uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6...

Method of increasing the deterrent to proliferation of nuclear fuels

DOEpatents

Rampolla, Donald S.

1982-01-01

A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

Galvanic cell for processing of used nuclear fuel

DOEpatents

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2017-02-07

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

Electrochemical fluorination for processing of used nuclear fuel

DOEpatents

Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

2016-07-05

A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors.

PubMed

Morreale, A C; Novog, D R; Luxat, J C

2012-01-01

Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU(®)(1) reactor allows for the potential production of large quantities of (99)Mo. This paper proposes (99)Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89-113% of the weekly world demand for (99)Mo can be obtained. Copyright © 2011 Elsevier Ltd. All rights reserved.

Data on Streamflow and Quality of Water and Bottom Sediment in and near Humboldt Wildlife Management Area, Churchill and Pershing Counties, Nevada, 1998-2000

USGS Publications Warehouse

Paul, Angela P.; Thodal, Carl E.

2003-01-01

This study was initiated to expand upon previous findings that indicated concentrations of dissolved solids, arsenic, boron, mercury, molybdenum, selenium, and uranium were either above geochemical background concentrations or were approaching or exceeding ecological criteria in the lower Humboldt River system. Data were collected from May 1998 to September 2000 to further characterize streamflow and surface-water and bottom-sediment quality in the lower Humboldt River, selected agricultural drains, Upper Humboldt Lake, and Lower Humboldt Drain (ephemeral outflow from Humboldt Sink). During this study, flow in the lower Humboldt River was either at or above average. Flows in Army and Toulon Drains generally were higher than reported in previous investigations. An unnamed agricultural drain contributed a small amount to the flow measured in Army Drain. In general, measured concentrations of sodium, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium were higher in water from agricultural drains than in Humboldt River water during this study. Mercury concentrations in water samples collected during the study period typically were below the laboratory reporting level. However, low-level mercury analyses showed that samples collected in August 1999 from Army Drain had higher mercury concentrations than those collected from the river or Toulon Drain or the Lower Humboldt Drain. Ecological criteria and effect concentrations for sodium, chloride, dissolved solids, arsenic, boron, mercury, and molybdenum were exceeded in some water samples collected as part of this study. Although water samples from the agricultural drains typically contained higher concentrations of sodium, chloride, dissolved solids, arsenic, boron, and uranium, greater instantaneous loads of these constituents were carried in the river near Lovelock than in agricultural drains during periods of high flow or non-irrigation. During this study, the high flows in the lower Humboldt River produced the maximum instantaneous loads of sodium, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium at all river-sampling sites, except molybdenum near Imlay. Nevada Division of Environmental Protection monitoring reports on mine-dewatering discharge for permitted releases of treated effluent to the surface waters of the Humboldt River and its tributaries were reviewed for reported discharges and trace-element concentrations from June 1998 to September 1999. These data were compared with similar information for the river near Imlay. In all bottom sediments collected for this study, arsenic concentrations exceeded the Canadian Freshwater Interim Sediment-Quality Guideline for the protection of aquatic life and probable-effect level (concentration). Sediments collected near Imlay, Rye Patch Reservoir, Lovelock, and from Toulon Drain and Army Drain were found to contain cadmium and chromium concentrations that exceeded Canadian criteria. Chromium concentrations in sediments collected from these sites also exceeded the consensus-based threshold-effect concentration. The Canadian criterion for sediment copper concentration was exceeded in sediments collected from the Humboldt River near Lovelock and from Toulon, Army, and the unnamed agricultural drains. Mercury in sediments collected near Imlay and from Toulon Drain in August 1999 exceeded the U.S. Department of the Interior sediment probable-effect level. Nickel concentrations in sediments collected during this study were above the consensus-based threshold-effect concentration. All other river and drain sediments had constituent concentrations below protective criteria and toxicity thresholds. In Upper Humboldt Lake, chloride, dissolved solids, arsenic, boron, molybdenum, and uranium concentrations in surface-water samples collected near the mouth of the Humboldt River generally were higher than in samples collected near the mouth of Army Drain. Ecological criteria or effect con

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

Joshi, Vineet V.; Paxton, Dean M.; Lavender, Curt A.

Over the past several years Pacific Northwest National Laboratory (PNNL) has been actively involved in supporting the U.S. Department of Energy National Nuclear Security Administration Office of Material Management and Minimization (formerly Global Threat Reduction Initiative). The U.S. High- Power Research Reactor (USHPRR) project is developing alternatives to existing highly enriched uranium alloy fuel to reduce the proliferation threat. One option for a high-density metal fuel is uranium alloyed with 10 wt% molybdenum (U-10Mo). Forming the U-10Mo fuel plates/foils via rolling is an effective technique and is actively being pursued as part of the baseline manufacturing process. The processing ofmore » these fuel plates requires systematic investigation/understanding of the pre- and post-rolling microstructure, end-state mechanical properties, residual stresses, and defects, their effect on the mill during processing, and eventually, their in-reactor performance. In the work documented herein, studies were conducted to determine the effect of cold and hot rolling the as-cast and homogenized U-10Mo on its microstructure and hardness. The samples were homogenized at 900°C for 48 h, then later annealed for several durations and temperatures to investigate the effect on the material’s microstructure and hardness. The rolling of the as-cast plate, both hot and cold, was observed to form a molybdenum-rich and -lean banded structure. The cold rolling was ineffective, and in some cases exacerbated the as-cast defects. The grains elongated along the rolling direction and formed a pancake shape, while the carbides fractured perpendicularly to the rolling direction and left porosity between fractured particles of UC. The subsequent annealing of these samples at sub-eutectoid temperatures led to rapid precipitation of the ' lamellar phase, mainly in the molybdenum-lean regions. Annealing the samples above the eutectoid temperature did not refine the grain size or the banded microstructure. However, annealing the samples led to quick recovery in hardness as evidenced by a drop in Vickers hardness of 20%. Hot rolling was performed at 650 and 800°C. The hot-rolling mill loads (load separation force) were approximately 40 to 50% less than the cold-rolling for the same reduction and thickness. It was observed that hot rolling the samples with 50% or more reduction in thickness were responsible for dynamic recrystallization in the hot-rolled samples and led to grain refinement. Unlike the cold-rolled samples, the hot-rolled samples did not fracture the carbides and appeared to heal the casting defects. The recovery phenomenon was similar to the cold-rolled samples above the eutectoid temperatures, but owing to the refined grain size, the precipitation of the lamellar phase was far more rapid in these samples and the hardness increased more rapidly than in the cold rolled sample when heated below the eutectoid temperature. The data generated from these rolling efforts has been used to make the process modeling efforts more robust and applicable to all USHPRR partner rolling mills. The flow stress for cold rolling the samples was determined to be between 170-190 ksi, with frictional forces between 0.2 and 0.4 for the PNNL mill. The measured roll separation forces and those simulated using finite element methods for hot and cold rolling for the PNNL rolling mill were in good agreement.« less

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2010 CFR

2010-07-01

... CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of effluent... uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6.71 pH (1) (1) 1...

40 CFR 471.71 - Effluent limitations representing the degree of effluent reduction attainable by the application...

Code of Federal Regulations, 2011 CFR

2011-07-01

... CATEGORY Uranium Forming Subcategory § 471.71 Effluent limitations representing the degree of effluent... uranium extruded Cadium 0.117 0.052 Chromium 0.152 0.062 Copper 0.654 0.344 Lead 0.145 0.069 Nickel 0.661 0.437 Fluoride 20.5 9.08 Molybdenum 2.28 1.18 Oil and grease 6.88 4.13 TSS 14.1 6.71 pH (1) (1) 1...

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

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

Tufic Madi Filho; Adonis Marcelo Saliba Silva; Jose Patricio Nahuel Cardenas

2015-07-01

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to createmore » samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)« less

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, J.P.; Miller, W.E.

1987-11-05

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

Geochemistry and migration of contaminants at the Weldon Spring chemical plant site, St. Charles County, Missouri, 1989-91

USGS Publications Warehouse

Schumacher, John G.

1993-01-01

The geochemistry of the shallow aquifer and geochemical controls on the migration of uranium and other constituents from raffinate pits were determined at the Weldon Spring chemical plant site. Surface-water samples from the raffinate pits con- tained large concentrations of calcium, magnesium, sodium, potassium, sulfate, nitrite, lithium, moly- bdenum, strontium, vanadium, and uranium. Analyses of interstitial-water samples from raffinate pit 3 indicated that concentrations of most constituents increased with increasing depth below the water- sediment interface. Nitrate and uranium were not chemically reduced and attenuated within the raffinate pits and can be expected to migrate into the overburden. Laboratory sorption experiments were performed to evaluate the effect of pH value on the sorption of several raffinate constituents by the overburden. No sorption of calcium, sodium, sulfate, nitrate, or lithium was observed. Sorption of molybdenum was dependent on solution pH and sorption of uranium was dependent on solution pH and carbonate concentration. The sorption of uranium and molybdenum was consistent with sorption controlled by oxyhydroxides. The quality of water collected in overburden lysimeters near raffinate pit 4 can be modeled as a mixture of water from raffinate pits 3 and 4, and an uncontaminated com- ponent in a system at equilibrium with ferrihydrite and calcite. Increased constituent concentrations in a perennial spring north of the site were the result of a subsurface connection between the spring and several losing stream segments receiving runoff from the site, in addition to seepage from the raffinate pits.

Continuous process electrorefiner

DOEpatents

Herceg, Joseph E [Naperville, IL; Saiveau, James G [Hickory Hills, IL; Krajtl, Lubomir [Woodridge, IL

2006-08-29

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

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-Mo/Si [41]). This analysis is furthermore in full agreement with the SEM/EDX characterisations which have highlighted the growth of a SiRDL in these U-Mo(Si)/Al_P fuel plates [30]. However it must be stressed that the amount of these U(Al,Si)3 and U3Si5 crystalline phases (about 0.3 wt%) is lower than the one obtained for fuel plates containing 4-6 wt% Si in the matrix [12]. It equals to the SiRDL amount measured in the IRIS4_2.1%Si fuel plate. Using these HE-XRD measurements, the Si concentration in SiRDLs is evaluated to 51 at%. This value is somewhat higher than when measured by EDX: it has been estimated to 40 at% in [30]. U2Mo and α"-U phase for compacts annealed at 340 °C, U2Mo and α'-U phase for compacts annealed at 450 °C [43], gamma;-U-Mo and α'-U for compacts annealed at 550 °C. These results obtained on compacts are in good agreement with previous works performed on U-8Mo ingots (see Fig. 9A) -even if some differences in the α-U phase structure must be mentioned - and in very close agreement with recent studies on thermally annealed U-Mo/Al fuel plates. Indeed destabilisation products found in this work are identical to those identified after fuel plate annealing at 550 °C [25] and 450 °C [43]. Moreover this work helps establishing that destabilisation products are U2Mo and α"-U at lower temperatures (below 450 °C). This was first demonstrated on fuel plates annealed at 425 °C for more than 50 h [43] and this is confirmed here with the analysis of the compacts annealed at 340 °C during 130 days. Note finally that whatever the presence of a coating, destabilisation ratios are very close in compacts annealed in the same conditions (see Fig. 9B) and that destabilisation ratios show the expected increase between 2 and 4 h annealing at 550 °C. The non-annealed U-Mo(Si)/Al compact has been lost during fabrication.

Resource potential for commodities in addition to Uranium in sandstone-hosted deposits: Chapter 13

USGS Publications Warehouse

Breit, George N.

2016-01-01

Sandstone-hosted deposits mined primarily for their uranium content also have been a source of vanadium and modest amounts of copper. Processing of these ores has also recovered small amounts of molybdenum, rhenium, rare earth elements, scandium, and selenium. These deposits share a generally common origin, but variations in the source of metals, composition of ore-forming solutions, and geologic history result in complex variability in deposit composition. This heterogeneity is evident regionally within the same host rock, as well as within districts. Future recovery of elements associated with uranium in these deposits will be strongly dependent on mining and ore-processing methods.

Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Edwards, Geoffrey W R; Priest, Nicholas D

2014-11-01

The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low as 1 mSv. In addition, if this method is extended so that Pu is also measured, then the combined amount of Pu and Pu is sufficiently high in the thorium-plutonium fuel that a committed effective dose of 1 mSv would be measurable. However, the fraction of Pu and Pu in the other two fuels is sufficiently low that a 1 mSv dose would remain below the detection limit using this technique. Thus new methods, such as fecal measurements of Pu (or other alpha emitters), will be required to measure exposure to these new fuels.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

DOEpatents

Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

1962-11-13

A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

An interdiffusional model for prediction of the interaction layer growth in the system uranium molybdenum/aluminum

NASA Astrophysics Data System (ADS)

Soba, A.; Denis, A.

2007-03-01

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U-Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U-Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U-Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

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

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

40 CFR 190.10 - Standards for normal operations.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Standards for the Uranium Fuel Cycle § 190.10 Standards for normal operations. Operations covered by this... radioactive materials, radon and its daughters excepted, to the general environment from uranium fuel cycle... the general environment from the entire uranium fuel cycle, per gigawatt-year of electrical energy...

77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-30

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice..., Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear...

Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

USGS Publications Warehouse

Finch, Warren Irvin

1997-01-01

The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

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

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.« less

Single-layer transition metal sulfide catalysts

DOEpatents

Thoma, Steven G [Albuquerque, NM

2011-05-31

Transition Metal Sulfides (TMS), such as molybdenum disulfide (MoS.sub.2), are the petroleum industry's "workhorse" catalysts for upgrading heavy petroleum feedstocks and removing sulfur, nitrogen and other pollutants from fuels. We have developed an improved synthesis technique to produce SLTMS catalysts, such as molybdenum disulfide, with potentially greater activity and specificity than those currently available. Applications for this technology include heavy feed upgrading, in-situ catalysis, bio-fuel conversion and coal liquefaction.

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Hazard evaluation of inorganics, singly and in mixtures, to Flannelmouth Sucker Catostomus latipinnis in the San Juan River, New Mexico

USGS Publications Warehouse

Hamilton, S.J.; Buhl, K.J.

1997-01-01

Larval flannelmouth sucker (Catostomus latipinnis) were exposed to arsenate, boron, copper, molybdenum, selenate, selenite, uranium, vanadium, and zinc singly, and to five mixtures of five to nine inorganics. The exposures were conducted in reconstituted water representative of the San Juan River near Shiprock, New Mexico. The mixtures simulated environmental ratios reported for sites along the San Juan River (San Juan River backwater, Fruitland marsh, Hogback East Drain, Mancos River, and McElmo Creek). The rank order of the individual inorganics, from most to least toxic, was: copper > zinc > vanadium > selenite > selenate > arsenate > uranium > boron > molybdenum. All five mixtures exhibited additive toxicity to flannelmouth sucker. In a limited number of tests, 44-day-old and 13-day-old larvae exhibited no difference in sensitivity to three mixtures. Copper was the major toxic component in four mixtures (San Juan backwater, Hogback East Drain, Mancos River, and McElmo Creek), whereas zinc was the major toxic component in the Fruitland marsh mixture, which did not contain copper. The Hogback East Drain was the most toxic mixture tested. Comparison of 96-h LC50values with reported environmental water concentrations from the San Juan River revealed low hazard ratios for arsenic, boron, molybdenum, selenate, selenite, uranium, and vanadium, moderate hazard ratios for zinc and the Fruitland marsh mixture, and high hazard ratios for copper at three sites and four environmental mixtures representing a San Juan backwater, Hogback East Drain, Mancos River, and McElmo Creek. The high hazard ratios suggest that inorganic contaminants could adversely affect larval flannelmouth sucker in the San Juan River at four sites receiving elevated inorganics.

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.; Smith, R. L.

1973-01-01

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

Comparison of abundances of chemical elements in mineralized and unmineralized sandstone of the Brushy Basin Member of the Morrison Formation, Smith Lake District, Grants uranium region, New Mexico

USGS Publications Warehouse

Pierson, C.T.; Spirakis, C.S.; Robertson, J.F.

1983-01-01

Statistical treatment of analytical data from the Mariano Lake and Ruby uranium deposits in the Smith Lake district, New Mexico, indicates that organic carbon, arsenic, barium, calcium, cobalt, copper, gallium, iron, lead, manganese, molybdenum, nickel, selenium, strontium, sulfur, vanadium, yttrium, and zirconium are concentrated along with uranium in primary ore. Comparison of the Smith Lake data with information from other primary deposits in the Grants uranium region and elsewhere in the Morrison Formation of the Colorado Plateau suggests that these elements, with the possible exceptions of zirconium and gallium and with the probable addition of aluminum and magnesium, are typically associated with primary, tabular uranium deposits. Chemical differences between the Ruby and Mariano Lake deposits are consistent with the interpretation that the Ruby deposit has been more affected by post-mineralization oxidizing solutions than has the Mariano Lake deposit.

Plasmachemical synthesis and evaluation of the thermal conductivity of metal-oxide compounds "Molybdenum-uranium dioxide"

NASA Astrophysics Data System (ADS)

Kotelnikova, Alexandra A.; Karengin, Alexander G.; Mendoza, Orlando

2018-03-01

The article represents possibility to apply oxidative and reducing plasma for plasma-chemical synthesis of metal-oxide compounds «Mo‒UO2» from water-salt mixtures «molybdic acid‒uranyl nitrate» and «molybdic acid‒ uranyl acetate». The composition of water-salt mixture was calculated and the conditions ensuring plasma-chemical synthesis of «Mo‒UO2» compounds were determined. Calculations were carried out at atmospheric pressure over a wide range of temperatures (300-4000 K), with the use of various plasma coolants (air, hydrogen). The heat conductivity coefficients of metal-oxide compounds «Mo‒UO2» consisting of continuous component (molybdenum matrix) are calculated. Inclusions from ceramics in the form of uranium dioxide were ordered in the matrix. Particular attention is paid to methods for calculating the coefficients of thermal conductivity of these compounds with the use of different models. Calculated results were compared with the experimental data.

DISSOLUTION OF URANIUM FUELS BY MONOOR DIFLUOROPHOSPHORIC ACID

DOEpatents

Johnson, R.; Horn, F.L.; Strickland, G.

1963-05-01

A method of dissolving and separating uranium from a uranium matrix fuel element by dissolving the uraniumcontaining matrix in monofluorophosphoric acid and/or difluorophosphoric acid at temperatures ranging from 150 to 275 un. Concent 85% C, thereafter neutralizing the solution to precipitate uranium solids, and converting the solids to uranium hexafluoride by treatment with a halogen trifluoride is presented. (AEC)

Sr 2Fe 1.5Mo 0.5O 6- δ as a regenerative anode for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Liu, Qiang; Bugaris, Daniel E.; Xiao, Guoliang; Chmara, Maxwell; Ma, Shuguo; zur Loye, Hans-Conrad; Amiridis, Michael D.; Chen, Fanglin

Sr 2Fe 1.5Mo 0.5O 6- δ (SFM) was prepared using a microwave-assisted combustion synthesis method. Rietveld refinement of powder X-ray diffraction data reveals that SFM crystallizes in the simple cubic perovskite structure with iron and molybdenum disordered on the B-site. No structure transition was observed by variable temperature powder X-ray diffraction measurements in the temperature range of 25-800 °C. XPS results show that the iron and molybdenum valences change with an increase in temperature, where the mixed oxidation states of both iron and molybdenum are believed to be responsible for the increase in the electrical conductivity with increasing temperature. SFM exhibits excellent redox stability and has been used as both anode and cathode for solid oxide fuel cells. Presence of sulfur species in the fuel or direct utilization of hydrocarbon fuel can result in loss of activity, however, as shown in this paper, the anode performance can be regenerated from sulfur poisoning or coking by treating the anode in an oxidizing atmosphere. Thus, SFM can be used as a regenerating anode for direct oxidation of sulfur-containing hydrocarbon fuels.

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-09-30

7 Uranium Mining and Milling ................................................................................................8...cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally occurring uranium lacks sufficient fissile 235U to make fuel for...enrichment, and finally fabrication into fuel elements. Australia exports its uranium after the mining and milling stage. Commercial enrichment services

Chemical state of fission products in irradiated uranium carbide fuel

NASA Astrophysics Data System (ADS)

Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

1987-12-01

The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

A physical model for evaluating uranium nitride specific heat

NASA Astrophysics Data System (ADS)

Baranov, V. G.; Devyatko, Yu. N.; Tenishev, A. V.; Khlunov, A. V.; Khomyakov, O. V.

2013-03-01

Nitride fuel is one of perspective materials for the nuclear industry. But unlike the oxide and carbide uranium and mixed uranium-plutonium fuel, the nitride fuel is less studied. The present article is devoted to the development of a model for calculating UN specific heat on the basis of phonon spectrum data within the solid state theory.

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

Control of a laser inertial confinement fusion-fission power plant

DOEpatents

Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

2015-10-27

A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

Fission gas bubble identification using MATLAB's image processing toolbox

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

Collette, R.

Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. This study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding proved to bemore » the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods. - Highlights: •Automated image processing can aid in the fuel qualification process. •Routines are developed to characterize fission gas bubbles in irradiated U–Mo fuel. •Frequency domain filtration effectively eliminates FIB curtaining artifacts. •Adaptive thresholding proved to be the most accurate segmentation method. •The techniques established are ready to be applied to large scale data extraction testing.« less

Nuclear fuel requirements for the American economy - A model

NASA Astrophysics Data System (ADS)

Curtis, Thomas Dexter

A model is provided to determine the amounts of various fuel streams required to supply energy from planned and projected nuclear plant operations, including new builds. Flexible, user-defined scenarios can be constructed with respect to energy requirements, choices of reactors and choices of fuels. The model includes interactive effects and extends through 2099. Outputs include energy provided by reactors, the number of reactors, and masses of natural Uranium and other fuels used. Energy demand, including electricity and hydrogen, is obtained from US DOE historical data and projections, along with other studies of potential hydrogen demand. An option to include other energy demand to nuclear power is included. Reactor types modeled include (thermal reactors) PWRs, BWRs and MHRs and (fast reactors) GFRs and SFRs. The MHRs (VHTRs), GFRs and SFRs are similar to those described in the 2002 DOE "Roadmap for Generation IV Nuclear Energy Systems." Fuel source choices include natural Uranium, self-recycled spent fuel, Plutonium from breeder reactors and existing stockpiles of surplus HEU, military Plutonium, LWR spent fuel and depleted Uranium. Other reactors and fuel sources can be added to the model. Fidelity checks of the model's results indicate good agreement with historical Uranium use and number of reactors, and with DOE projections. The model supports conclusions that substantial use of natural Uranium will likely continue to the end of the 21st century, though legacy spent fuel and depleted uranium could easily supply all nuclear energy demand by shifting to predominant use of fast reactors.

Chemical vapor deposition of Mo tubes for fuel cladding applications

DOE PAGES

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.; ...

2018-01-31

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Chemical vapor deposition of Mo tubes for fuel cladding applications

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

Beaux, Miles F.; Vodnik, Douglas R.; Peterson, Reuben J.

In this study, chemical vapor deposition (CVD) techniques have been evaluated for fabrication of free-standing 0.25 mm thick molybdenum tubes with the end goal of nuclear fuel cladding applications. In order to produce tubes with the wall thickness and microstructures desirable for this application, long deposition durations on the order of 50 h with slow deposition rates were employed. A standard CVD method, involving molybdenum pentachloride reduction by hydrogen, as well as a fluidized-bed CVD (FBCVD) method was applied towards these objectives. Characterization of the tubes produced in this manner revealed regions of material with fine grain microstructure and wallmore » thickness suitable for fuel cladding applications, but lacking necessary uniformity across the length of the tubes. Finally, a path forward for the production of freestanding molybdenum tubes that possess the desired properties across their entire length has been identified and can be accomplished by future optimization of the deposition system.« less

Uranium production in Romania

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

NONE

1994-12-01

This article reviews uranium production in Romania. Geological aspects of the country are discussed, and known uranium deposits are noted. Uranium mining and milling activities are also covered. Utilization of Romania`s uranium production industry will primarily be to supply the country`s nuclear power program, and with the present adequate supplies and the operation of their recently revamped fuel production facility, Romania should be self-reliant in the front end of the nuclear fuel cycle.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

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

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no usemore » of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.« less

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

NUCLEAR FUEL COMPOSITION

DOEpatents

Spedding, F.H.; Wilhelm, H.A.

1960-05-31

A novel reactor composition for use in a self-sustaining fast nuclear reactor is described. More particularly, a fuel alloy comprising thorium and uranium-235 is de scribed, the uranium-235 existing in approximately the same amount that it is found in natural uranium, i.e., 1.4%.

Molybdenum Availability Is Key to Nitrate Removal in Contaminated Groundwater Environments

PubMed Central

Thorgersen, Michael P.; Lancaster, W. Andrew; Vaccaro, Brian J.; Poole, Farris L.; Rocha, Andrea M.; Mehlhorn, Tonia; Pettenato, Angelica; Ray, Jayashree; Waters, R. Jordan; Melnyk, Ryan A.; Chakraborty, Romy; Deutschbauer, Adam M.; Arkin, Adam P.

2015-01-01

The concentrations of molybdenum (Mo) and 25 other metals were measured in groundwater samples from 80 wells on the Oak Ridge Reservation (ORR) (Oak Ridge, TN), many of which are contaminated with nitrate, as well as uranium and various other metals. The concentrations of nitrate and uranium were in the ranges of 0.1 μM to 230 mM and <0.2 nM to 580 μM, respectively. Almost all metals examined had significantly greater median concentrations in a subset of wells that were highly contaminated with uranium (≥126 nM). They included cadmium, manganese, and cobalt, which were 1,300- to 2,700-fold higher. A notable exception, however, was Mo, which had a lower median concentration in the uranium-contaminated wells. This is significant, because Mo is essential in the dissimilatory nitrate reduction branch of the global nitrogen cycle. It is required at the catalytic site of nitrate reductase, the enzyme that reduces nitrate to nitrite. Moreover, more than 85% of the groundwater samples contained less than 10 nM Mo, whereas concentrations of 10 to 100 nM Mo were required for efficient growth by nitrate reduction for two Pseudomonas strains isolated from ORR wells and by a model denitrifier, Pseudomonas stutzeri RCH2. Higher concentrations of Mo tended to inhibit the growth of these strains due to the accumulation of toxic concentrations of nitrite, and this effect was exacerbated at high nitrate concentrations. The relevance of these results to a Mo-based nitrate removal strategy and the potential community-driving role that Mo plays in contaminated environments are discussed. PMID:25979890

Ground-water contamination near a uranium tailings disposal site in Colorado

USGS Publications Warehouse

Goode, Daniel J.; Wilder, Russell J.

1987-01-01

Contaminants from uranium tailings disposed of at an active mill in Colorado have seeped into the shallow ground water onsite. This ground water discharges into the Arkansas River Valley through a superposed stream channel cut in the resistant sandstone ridge at the edge of a synclinal basin. In the river valley, seasonal surface-water irrigation has a significant impact on hydrodynamics. Water levels in residential wells fluctuate up to 20 ft and concentrations of uranium, molybdenum, and other contaminants also vary seasonally, with highest concentrations in the Spring, prior to irrigation, and lowest concentrations in the Fall. Results of a simple transient mixing cell model support the hypothesis that lateral ground-water inflow, and not irrigation recharge, is the source of ground-water contamination.

ALD coating of nuclear fuel actinides materials

DOEpatents

Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

2017-09-05

The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

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.

A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors

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

Martinez-Frances, N.; Timm, W.; Rossbach, D.

2012-07-01

Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main designmore » criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

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

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Uranium extraction: Fuel from seawater

DOE PAGES

Tsouris, Costas; Oak Ridge National Lab.

2017-02-17

Over four billion tonnes of uranium are currently in the oceans that could be harvested for nuclear fuel, but current capture methods have limited performance and reusability. Now, an electrochemical method using modified carbon electrodes is shown to be promising for the extraction of uranium from seawater.

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

Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Ratavia, IL

2007-09-11

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

Porous membrane electrochemical cell for uranium and transuranic recovery from molten salt electrolyte

DOEpatents

Willit, James L [Batavia, IL

2010-09-21

An improved process and device for the recovery of the minor actinides and the transuranic elements (TRU's) from a molten salt electrolyte. The process involves placing the device, an electrically non-conducting barrier between an anode salt and a cathode salt. The porous barrier allows uranium to diffuse between the anode and cathode, yet slows the diffusion of uranium ions so as to cause depletion of uranium ions in the catholyte. This allows for the eventual preferential deposition of transuranics present in spent nuclear fuel such as Np, Pu, Am, Cm. The device also comprises an uranium oxidation anode. The oxidation anode is solid uranium metal in the form of spent nuclear fuel. The spent fuel is placed in a ferric metal anode basket which serves as the electrical lead or contact between the molten electrolyte and the anodic uranium metal.

75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-29

... Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services, National... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and...

JACKETED FUEL ELEMENT

DOEpatents

Wigner, E.P.; Szilard, L.; Creutz, E.C.

1959-02-01

These fuel elements are comprised of a homogeneous metallic uranium body completely enclosed and sealed in an aluminum cover. The uranium body and aluminum cover are bonded together by a layer of zinc located between them. The bonding layer serves to improve transfer of heat, provides an additional protection against corrosion of the uranium by the coolant, and also localizes any possible corrosion by preventing travel of corrosive material along the surface of the fuel element.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

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

Hammon, Duncan L.; Clarke, Kester D.; Alexander, David J.

2015-05-29

The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield.more » This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.« less

Rolling Process Modeling Report: Finite-Element Prediction of Roll Separating Force and Rolling Defects

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

Soulami, Ayoub; Lavender, Curt A.; Paxton, Dean M.

2014-04-23

Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum (U-10Mo) alloy plate-type fuel for the U.S. high-performance research reactors. This work supports the Convert Program of the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Global Threat Reduction Initiative. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll separating forces and rolling defects. Simulations were performed using a finite-element model developed using the commercial code LS-Dyna. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel have been conducted following two different schedules. Model predictions ofmore » the roll-separation force and roll-pack thicknesses at different stages of the rolling process were compared with experimental measurements. This report discusses various attributes of the rolled coupons revealed by the model (e.g., dog-boning and thickness non-uniformity).« less

CATALYTIC RECOMBINATION OF RADIOLYTIC GASES IN THORIUM OXIDE SLURRIES

DOEpatents

Morse, L.E.

1962-08-01

A method for the coinbination of hydrogen and oxygen in aqueous thorium oxide-uranium oxide slurries is described. A small amount of molybdenum oxide catalyst is provided in the slurry. This catalyst is applicable to the recombination of hydrogen and/or deuterium and oxygen produced by irradiation of the slurries in nuclear reactors. (AEC)

Strength and fracture of uranium, plutonium and several their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, V. K.

2012-08-01

Results on studying the spall fracture of uranium, plutonium and several their alloys under shock wave loading are presented in the paper. The problems of influence of initial temperature in a range of - 196 - 800∘C and loading time on the spall strength and failure character of uranium and two its alloys with molybdenum and both molybdenum and zirconium were studied. The results for plutonium and its alloy with gallium were obtained at a normal temperature and in a temperature range of 40-315∘C, respectively. The majority of tests were conducted with the samples in the form of disks 4 mm in thickness. They were loaded by the impact of aluminum plates 4 mm thick through a copper screen 12 mm thick serving as the cover or bottom part of a special container. The character of spall failure of materials and the damage degree of samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. The conditions of shock wave loading were calculated using an elastic-plastic computer program. The comparison of obtained results with the data of other researchers on the spall fracture of examined materials was conducted.

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.

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-12-01

Enrichment.......................................................................................................7 Uranium Mining and Milling...Issues for Congress Congressional Research Service 7 The nuclear fuel cycle begins with mining uranium ore and upgrading it to yellowcake. Because...uranium after the mining and milling stage. Commercial enrichment services are available in the United States, Europe, Russia, and Japan. Fuel

Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

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

O'Donnell, F.R.; Hoy, H.C.

1981-10-01

Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/supmore » 12/ Btu of energy output, and per other appropriate units of output.« less

Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

DOE PAGES

George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; ...

2014-12-01

Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO 2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO 2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Handwerk, J.H.; BAch, R.A.

1959-08-18

A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

Neutronics qualification of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program - Transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

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

Leray, O.; Hudelot, J. P.; Antony, M.

2011-07-01

The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in {sup 235}U) fuels (U{sub 3}Si{sub 2} for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has beenmore » performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl{sub x}/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm{sup 3}. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% {sup 235}U) by the Minerve-dedicated interpretation tool: PIMS. The effect of energy meshes and evaluations put forward the JEFF3.1.1/SHEM scheme that leads to a better calculation of the reactivity effect of VALMONT samples. Then, in order to quantify the impact of the uncertainties linked to the basic nuclear data, their propagation from the cross section measurement to the final computational result was analysed in a rigorous way by using a nuclear data re-estimation method based on Gauss-Newton iterations. This study concludes that the prior uncertainties due to nuclear data (uranium, aluminium, beryllium and water) on the reactivity of the Begin Of Cycle (BOC) for the JHR core reach 1217 pcm at 2{sigma}. Now, the uppermost uncertainty on the JHR reactivity is due to aluminium. (authors)« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

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

Beeler, Benjamin; Baskes, Michael; Andersson, David

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

DOE PAGES

Beeler, Benjamin; Baskes, Michael; Andersson, David; ...

2017-08-18

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel bene ts from higher thermal conductivity and higher ssile density compared to uranium dioxide (UO 2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling e orts are underway to address this gap in knowledge. In this study, a semi-empirical modi ed Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is ttedmore » to the formation energy, defect energies and structural properties of U 3Si 2. The primary phase of interest (U 3Si 2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.« less

A modified Embedded-Atom Method interatomic potential for uranium-silicide

NASA Astrophysics Data System (ADS)

Beeler, Benjamin; Baskes, Michael; Andersson, David; Cooper, Michael W. D.; Zhang, Yongfeng

2017-11-01

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential is fitted to the formation energy, defect energies and structural properties of U3Si2. The primary phase of interest (U3Si2) is accurately described over a wide temperature range and displays good behavior under irradiation and with free surfaces. The potential can also describe a variety of U-Si phases across the composition spectrum.

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 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.

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.

Grain boundary engineering to control the discontinuous precipitation in multicomponent U10Mo alloy

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

Devaraj, Arun; Kovarik, Libor; Kautz, Elizabeth

Grain boundaries in metallic alloys often play a crucial role, not only in determining the mechanical properties or thermal stability of alloys, but also in dictating the phase transformation kinetics during thermomechanical processing. We demonstrate that locally stabilized structure and compositional segregation at grain boundaries—“grain boundary complexions”—in a complex multicomponent alloy can be modified to influence the kinetics of cellular transformation during subsequent thermomechanical processing. Using aberration-corrected scanning transmission electron microscopy and atom probe tomography analysis of a metallic nuclear fuel highly relevant to worldwide nuclear non-proliferation efforts —uranium-10 wt% molybdenum (U-10Mo) alloy, new evidence for the existence of grainmore » boundary complexion is provided. We then modified the concentration of impurities dissolved in Υ-UMo grain interiors and/or segregated to Υ-UMo grain boundaries by changing the homogenization treatment, and these effects were used used to retard the kinetics of cellular transformation during subsequent sub-eutectoid annealing in this U-10-Mo alloy during sub-eutectoid annealing. Thus, this work provided insights on tailoring the final microstructure of the U-10Mo alloy, which can potentially improve the irradiation performance of this important class of alloy fuels.« less

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, D.J.; McTaggart, D.R.

1983-08-31

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, David J.; McTaggart, Donald R.

1984-01-01

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

U.sup.+4 generation in HTER

DOEpatents

Miller, William E [Naperville, IL; Gay, Eddie C [Park Forest, IL; Tomczuk, Zygmunt [Homer Glen, IL

2006-03-14

A improved device and process for recycling spent nuclear fuels, in particular uranium metal, that facilitates the refinement and recovery of uranium metal from spent metallic nuclear fuels. The electrorefiner device comprises two anodes in predetermined spatial relation to a cathode. The anodese have separate current and voltage controls. A much higher voltage than normal for the electrorefining process is applied to the second anode, thereby facilitating oxidization of uranium (III), U.sup.+, to uranium (IV), U.sup.+4. The current path from the second anode to the cathode is physically shorter than the similar current path from the second anode to the spent nuclear fuel contained in a first anode shaped as a basket. The resulting U.sup.+4 oxidizes and solubilizes rough uranium deposited on the surface of the cathode. A softer uranium metal surface is left on the cathode and is more readily removed by a scraper.

Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

NASA Astrophysics Data System (ADS)

Soulami, Ayoub; Burkes, Douglas E.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean

2017-10-01

A major goal of the Convert Program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) is to enable high-performance research reactors to operate with low-enriched uranium rather than the high-enriched uranium currently used. To this end, uranium alloyed with 10 wt% molybdenum (U-10Mo) represents an ideal candidate because of its stable gamma phase, low neutron caption cross section, acceptable swelling response, and predictable irradiation behavior. However, because of the complexities of the fuel design and the need for rolled monolithic U-10Mo foils, new developments in processing and fabrication are necessary. This study used a finite-element code, LS-DYNA, as a predictive tool to optimize the rolling process. Simulations of the hot rolling of U-10Mo coupons encapsulated in low-carbon steel were conducted following two different schedules. Model predictions of the roll-separation force and roll pack thicknesses at different stages of the rolling process were compared with experimental measurements. The study reported here discussed various attributes of the rolled coupons revealed by the model (e.g., waviness and thickness non-uniformity like dog-boning). To investigate the influence of the cladding material on these rolling defects, other cases were simulated: hot rolling with alternative can materials, namely, 304 stainless steel and Zircaloy-2, and bare-rolling. Simulation results demonstrated that reducing the mismatch in strength between the coupon and can material improves the quality of the rolled sheet. Bare-rolling simulation results showed a defect-free rolled coupon. The finite-element model developed and presented in this study can be used to conduct parametric studies of several process parameters (e.g., rolling speed, roll diameter, can material, and reduction).

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

DOEpatents

Tomczuk, Zygmunt; Miller, William E.; Wolson, Raymond D.; Gay, Eddie C.

1991-01-01

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... uranium or enriching uranium in the isotope 235, zirconium tubes, heavy water or deuterium, nuclear-grade..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

Spark plasma sintering and microstructural analysis of pure and Mo doped U3Si2 pellets

NASA Astrophysics Data System (ADS)

Lopes, Denise Adorno; Benarosch, Anna; Middleburgh, Simon; Johnson, Kyle D.

2017-12-01

U3Si2 has been considered as an alternative fuel for Light Water Reactors (LWRs) within the Accident Tolerant Fuels (ATF) initiative, begun after the Fukushima-Daiichi Nuclear accidents. Its main advantages are high thermal conductivity and high heavy metal density. Despite these benefits, U3Si2 presents an anisotropic crystallographic structure and low solubility of fission products, which can result in undesirable effects under irradiation conditions. In this paper, spark plasma sintering (SPS) of U3Si2 pellets is studied, with evaluation of the resulting microstructure. Additionally, exploiting the short sintering time in SPS, a molybdenum doped pellet was produced to investigate the early stages of the Mo-U3Si2 interaction, and analyze how this fission product is accommodated in the fuel matrix. The results show that pellets of U3Si2 with high density (>95% TD) can be obtained with SPS in the temperature range of 1200°C-1300 °C. Moreover, the short time employed in this technique was found to generate a unique microstructure for this fuel, composed mainly of closed nano-pores (<1 μm) and small average grain size (∼4.5 μm). The addition of Mo (1.5 at%) demonstrated no solubility of Mo in the U3Si2 matrix. The interaction of this fission product with the fuel matrix at 1200 °C formed, in the early stages, the stoichiometric U2Mo3Si4 ternary as well as precipitates of free uranium with small quantities of dissolved Si and Mo at the front of the reaction.

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

NASA Astrophysics Data System (ADS)

Leenaers, A.; Detavernier, C.; Van den Berghe, S.

2008-11-01

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si) 3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si) 3 layer.

Molybdenum Availability Is Key to Nitrate Removal in Contaminated Groundwater Environments

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

Thorgersen, Michael P.; Lancaster, W. Andrew; Vaccaro, Brian J.

2015-05-15

The concentrations of molybdenum (Mo) and 25 other metals were measured in groundwater samples from 80 wells on the Oak Ridge Reservation (ORR) (Oak Ridge, TN), many of which are contaminated with nitrate, as well as uranium and various other metals. Moreover, the concentrations of nitrate and uranium were in the ranges of 0.1 μM to 230 mM and <0.2 nM to 580 μM, respectively. Most metals examined had significantly greater median concentrations in a subset of wells that were highly contaminated with uranium (≥126 nM). They included cadmium, manganese, and cobalt, which were 1,300- to 2,700-fold higher. A notablemore » exception, however, was Mo, which had a lower median concentration in the uranium-contaminated wells. This is significant, because Mo is essential in the dissimilatory nitrate reduction branch of the global nitrogen cycle. It is required at the catalytic site of nitrate reductase, the enzyme that reduces nitrate to nitrite. Furthermore, more than 85% of the groundwater samples contained less than 10 nM Mo, whereas concentrations of 10 to 100 nM Mo were required for efficient growth by nitrate reduction for twoPseudomonasstrains isolated from ORR wells and by a model denitrifier,Pseudomonas stutzeriRCH2. Higher concentrations of Mo tended to inhibit the growth of these strains due to the accumulation of toxic concentrations of nitrite, and this effect was exacerbated at high nitrate concentrations. The relevance of these results to a Mo-based nitrate removal strategy and the potential community-driving role that Mo plays in contaminated environments are discussed.« less

THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS

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

Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.

2011-07-17

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies requiredmore » to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.« less

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2017-12-01

It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-18

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material Safety, and Safeguards. [FR Doc...

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

PubMed

Tendall, Danielle M; Binder, Claudia R

2011-03-15

The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

Depleted uranium startup of spent-fuel treatment operations at ANL-West

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

Goff, K.M.; Mariani, R.D.; Bonomo, N.L.

1995-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of Experimental Breeder Reactor II (EBR-II) spent nuclear fuel. This fuel will be treated using an electrometallurgical process in the fuel conditioning facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The process equipment is undergoing testing with depleted uranium in preparation for irradiated fuel operations during the summer of 1995.

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

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

This supplemental report describes fuel fabrication efforts conducted for the Idaho National Laboratory Trade Study for the TREAT Conversion project that is exploring the replacement of the HEU (Highly Enriched Uranium) fuel core of the TREAT reactor with LEU (Low Enriched Uranium) fuel. Previous reports have documented fabrication of fuel by the “upgrade” process developed at Los Alamos National Laboratory. These experiments supplement an earlier report that describes efforts to increase the graphite content of extruded fuel and minimize cracking.

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

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

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

DOEpatents

Sease, J.D.; Harrington, F.E.

1973-12-11

Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

Evaluation of Non-Oxide Fuel for Fission-based Nuclear Reactors on Spacecraft

DTIC Science & Technology

smaller and potentially lighter core, whichis a significant advantage. The results of this study indicate that use of both UC and UN may result in significant weight savings due tohigher uranium loading density....The goal of this project was to study the performance of atypical uranium-based fuels in a nuclear reactor capable of producing 1 megawattof thermal...UN), or uranium carbide (UC) and compared their performance to uranium oxide (UO2) which is thefuel form used in the vast majority of commercial

Supply of enriched uranium for research reactors

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

Mueller, H.

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

PubMed

Rao, Ankita; Kumar Sharma, Abhishek; Kumar, Pradeep; Charyulu, M M; Tomar, B S; Ramakumar, K L

2014-07-01

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards. Copyright © 2014 Elsevier Ltd. All rights reserved.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Technical Reports Server (NTRS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

1987-01-01

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1962-04-10

A process of recovering plutonium from neutronbombarded uranium fuel by dissolving the fuel in equimolar aluminum chloride-potassium chloride; heating the mass to above 700 deg C for decomposition of plutonium tetrachloride to the trichloride; extracting the plutonium trichloride into a molten salt containing from 40 to 60 mole % of lithium chloride, from 15 to 40 mole % of sodium chloride, and from 0 to 40 mole % of potassium chloride or calcium chloride; and separating the layer of equimolar chlorides containing the uranium from the layer formed of the plutonium-containing salt is described. (AEC)

Uranium nitride fuel fabrication for SP-100 reactors

NASA Astrophysics Data System (ADS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

NASA Technical Reports Server (NTRS)

Bowles, K. J.; Gluyas, R. E.

1975-01-01

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

The Use of Thorium within the Nuclear Power Industry - 13472

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

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-10-01

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

U-10Mo Sample Preparation and Examination using Optical and Scanning Electron Microscopy

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

Prabhakaran, Ramprashad; Joshi, Vineet V.; Rhodes, Mark A.

2016-03-30

The purpose of this document is to provide guidelines to prepare specimens of uranium alloyed with 10 weight percent molybdenum (U-10Mo) for optical metallography and scanning electron microscopy. This document also provides instructions to set up an optical microscope and a scanning electron microscope to analyze U-10Mo specimens and to obtain the required information.

Equilibrium of molybdenum in selected extraction systems

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

Tkac, Peter; Paulenova, Alena

2007-07-01

The concentration of molybdenum(VI) in dissolved irradiated nuclear fuel is comparable with the concentrations of Tc, Am and Np. Therefore it is of big interest to understand its behavior under conditions related to the UREX/TRUEX process. The effect of the poly-speciation of molybdenum in aqueous solution on its extraction by neutral solvents TBP and CMPO/TBP was studied. Extraction yields of molybdenum decreased significantly when AHA was added to aqueous phase. Our investigation confirmed a strong ability of the aceto-hydroxamic acid to form complexes with Mo in high acidic solutions. Spectroscopic data (UV-Vis) confirmed that a fraction of the Mo(VI)-AHA complexmore » is present in the organic phase after extraction. (authors)« less

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

Moeller, Helen H.; Bremser, Albert H.; Gontar, Alexander; Fiviesky, Evgeny

1997-01-01

Specialists in space applications are currently focusing on bimodal power systems designed to provide both electric power and thermal propulsion (Kennedy, 1994 and Houts, 1995). Our work showed that thermionics is a viable technology for nuclear bimodal power systems. We demonstrated that materials for a thermionic fuel-emitter combination capable of performing at operating temperatures of 2473 K are not only possible but available. The objective of this work, funded by the US Department of Energy, Office of Space and Defense Power Systems, was to evaluate the compatibility of fuel material consisting of an uranium carbide/tantalum carbide solid solution with an emitter material consisting of a monocrystalline tungsten-niobium alloy. The uranium loading of the fuel material was 70 mole% uranium carbide. The program was successfully accomplished by a B&W/SIA LUTCH team. Its workscope was integrated with tasks being performed at both Babcock & Wilcox, Lynchburg Research Center, Lynchburg, Virginia, and SIA LUTCH, Podolsk, Russia. Samples were fabricated by LUTCH and seven thermal tests were performed in a hydrogen atmosphere. The first preliminary test was performed at 2273 K by LUTCH, and the remaining six tests were performed At B&W. Three tests were performed at 2273 K, two at 2373 K, and the final test at 2473 K. The results showed that the fuel and emitter materials were compatible in the presence of hydrogen. No evidence of liquid formation, dissolution of the uranium carbide from the uranium carbide/tantalum carbide solid solution, or diffusion of the uranium into the monocrystalline tungsten alloy was observed. Among the highlights of the program was the successful export of the fuel samples from Russia and their import into the US by commercial transport. This paper will discuss the technical aspects of this work.

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

NASA Technical Reports Server (NTRS)

Rohal, R. G.; Tambling, T. N.

1973-01-01

Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

Letter Report: Looking Ahead at Nuclear Fuel Resources

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

J. Stephen Herring

2013-09-01

The future of nuclear energy and its ability to fulfill part of the world’s energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energymore » community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earth’s crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials’ locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.« less

Loading blended, low-enriched uranium fuel in browns ferry units 2 and 3

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

Brown, C.; Eichenberg, T.; Haun, J.

2006-07-01

This paper summarizes fuel and cycle design results for the Tennessee Valley Authority (TVA) / Dept. of Energy (DOE) program to burn blended, low-enriched uranium (BLEU) material in the Browns Ferry Nuclear Units 2 and 3. The BLEU material typically has about 60 times the allowed limit of U-236 in what would be defined as commercial, i.e., virgin, uranium. U-236 in particular is a strong neutron absorber. Also included is a comparison of cycles using commercial uranium versus BLEU to determine the impact on key core design parameters of the high U-236 content in the BLEU. Finally, there is amore » short discussion of the economic advantages of BLEU fuel. (authors)« less

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.

Preparation of uranium fuel kernels with silicon carbide nanoparticles using the internal gelation process

NASA Astrophysics Data System (ADS)

Hunt, R. D.; Silva, G. W. C. M.; Lindemer, T. B.; Anderson, K. K.; Collins, J. L.

2012-08-01

The US Department of Energy continues to use the internal gelation process in its preparation of tristructural isotropic coated fuel particles. The focus of this work is to develop uranium fuel kernels with adequately dispersed silicon carbide (SiC) nanoparticles, high crush strengths, uniform particle diameter, and good sphericity. During irradiation to high burnup, the SiC in the uranium kernels will serve as getters for excess oxygen and help control the oxygen potential in order to minimize the potential for kernel migration. The hardness of SiC required modifications to the gelation system that was used to make uranium kernels. Suitable processing conditions and potential equipment changes were identified so that the SiC could be homogeneously dispersed in gel spheres. Finally, dilute hydrogen rather than argon should be used to sinter the uranium kernels with SiC.

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

Argonne explains nuclear recycling in 4 minutes

ScienceCinema

Willit, Jim; Williamson, Mark; Haynes, Amber

2018-05-30

Currently, when using nuclear energy only about five percent of the uranium used in a fuel rod gets fissioned for energy; after that, the rods are taken out of the reactor and put into permanent storage. There is a way, however, to use almost all of the uranium in a fuel rod. Recycling used nuclear fuel could produce hundreds of years of energy from just the uranium we've already mined, all of it carbon-free. Problems with older technology put a halt to recycling used nuclear fuel in the United States, but new techniques developed by scientists at Argonne National Laboratory address many of those issues. For more information, visit http://www.anl.gov/energy/nuclear-energy.

NUCLEAR REACTOR FUEL SYSTEMS

DOEpatents

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

The Permo-Triassic uranium deposits of Gondwanaland

NASA Astrophysics Data System (ADS)

le Roux, J. P.; Toens, P. D.

The world's uranium provinces are time bound and occur in five distinct periods ranging from the Proterozoic to the Recent. One of these periods embraces the time of Gondwana sedimentation and probably is related to the proliferation of land plants from the Devonian on-ward. Decaying vegetal matter produced reducing conditions that enhanced uranium precipitation. The association of uranium with molassic basins adjacent to uplifted granitic and volcanic arcs suggests that lithospheric plate subduction, leading to anatexis of basement rocks and andesitic volcanism, created favorable conditions for uranium mineralization. Uranium occurrences of Gondwana age are of four main types: sandstone-hosted, coal-hosted, pelite-hosted, and vein-type deposits. Sandstone-hosted deposits commonly occur in fluviodeltaic sediments and are related to the presence of organic matter. These deposits commonly are enriched in molybdenum and other base metal sulfides and have been found in South Africa, Zimbabwe, Zambia, Angola, Niger, Madagascar, India, Australia, Argentina, and Brazil. Coalhosted deposits contain large reserves of uranium but are of low grade. In Africa they are mostly within the Permian Ecca Group and its lateral equivalents, as in the Springbok Flats, Limpopo, Botswana, and Tanzania basins. Uraniferous black shales are present in the Gabon and Amazon basins but grades are low. Vein-type uranium is found in Argentina, where it occurs in clustered veins crosscutting sedimentary rocks and quartz porphyries.

National Uranium Resource Evaluation Program. Hydrogeochemical and stream sediment reconnaissance basic data for Beeville NTMS Quadrangle, Texas. Uranium resource evaluation project

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

Not Available

Results of a reconnaissance geochemical survey of the Beeville Quadrangle, Texas are reported. Field and laboratory data are presented for 373 groundwater and 364 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are displayed. A generalized geologic map of the survey area is provided, and pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. The groundwater data indicate that the northwestern corner of the quadrangle is the most favorable for potential uranium mineralization. Favorability is indicated by high uranium concentrations; high arsenic, molybdenum, and vanadium concentrations;more » and proximity and similar geologic setting to the mines of the Karnes County mining district. Other areas that appear favorable are an area in Bee and Refugio Counties and the northeastern part of the quadrangle. Both areas have water chemistry similar to the Karnes County area, but the northeastern area does not have high concentrations of pathfinder elements. The stream sediment data indicate that the northeastern corner of the quadrangle is the most favorable for potential mineralization, but agricultural practices and mineralogy of the outcropping Beaumont Formation may indicate a false anomaly. The northwestern corner of the quadrangle is considered favorable because of its proximity to the known uranium deposits, but the data do not seem to support this.« less

Uranium from German Nuclear Power Projects of the 1940s— A Nuclear Forensic Investigation

PubMed Central

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-01-01

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel.3b,d, 4 Through measurement of the 230Th/234U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the 87Sr/86Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of 236U and 239Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. PMID:26501922

Uranium dioxide electrolysis

DOEpatents

Willit, James L [Batavia, IL; Ackerman, John P [Prescott, AZ; Williamson, Mark A [Naperville, IL

2009-12-29

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

Returning HEU Fuel from the Czech Republic to Russia

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

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Zirconium determination by cooling curve analysis during the pyroprocessing of used nuclear fuel

NASA Astrophysics Data System (ADS)

Westphal, B. R.; Price, J. C.; Bateman, K. J.; Marsden, K. C.

2015-02-01

An alternative method to sampling and chemical analyses has been developed to monitor the concentration of zirconium in real-time during the casting of uranium products from the pyroprocessing of used nuclear fuel. The method utilizes the solidification characteristics of the uranium products to determine zirconium levels based on standard cooling curve analyses and established binary phase diagram data. Numerous uranium products have been analyzed for their zirconium content and compared against measured zirconium data. From this data, the following equation was derived for the zirconium content of uranium products:

Method of Making Uranium Dioxide Bodies

DOEpatents

Wilhelm, H. A.; McClusky, J. K.

1973-09-25

Sintered uranium dioxide bodies having controlled density are produced from U.sub.3 O.sub.8 and carbon by varying the mole ratio of carbon to U.sub.3 O.sub.8 in the mixture, which is compressed and sintered in a neutral or slightly oxidizing atmosphere to form dense slightly hyperstoichiometric uranium dioxide bodies. If the bodies are to be used as nuclear reactor fuel, they are subsequently heated in a hydrogen atmosphere to achieve stoichiometry. This method can also be used to produce fuel elements of uranium dioxide -- plutonium dioxide having controlled density.

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

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

Yang, Won Sik; Lin, C. S.; Hader, J. S.

2016-01-30

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the dischargedmore » fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements« 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

DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

DOEpatents

Zegler, S.T.

1963-11-01

The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

Measurement of thermal diffusivity of depleted uranium metal microspheres

NASA Astrophysics Data System (ADS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

10 CFR 51.75 - Draft environmental impact statement-construction permit, early site permit, or combined licesne.

Code of Federal Regulations, 2010 CFR

2010-01-01

... storage of spent fuel for the nuclear power plant within the scope of the generic determination in § 51.23..., and 51.73. The contribution of the environmental effects of the uranium fuel cycle activities....71, 51.72, 51.73, and this section. The contribution of the environmental effects of the uranium fuel...

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

Significance of breeding in fast nuclear reactors

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

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, A.

1985-10-25

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

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

Jiang, J.; Yuan, B.; Jin, M.

2012-07-01

Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

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

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

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

James A. Smith; Barry H. Rabin; Mathieu Perton

2012-07-01

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process.more » Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.« less

Laser shockwave technique for characterization of nuclear fuel plate interfaces

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

Perton, M.; Levesque, D.; Monchalin, J.-P.

2013-01-25

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process.more » Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.« less

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

A method of casting a protective jacket about a ura nium fuel element that will bond completely to the uranium without the use of stringers or supports that would ordinarily produce gaps in the cast metal coating and bond is presented. Preformed endcaps of alumlnum alloyed with 13% silicon are placed on the ends of the uranium fuel element. These caps will support the fuel element when placed in a mold. The mold is kept at a ing alloy but below that of uranium so the cast metal jacket will fuse with the endcaps forming a complete covering and bond to the fuel element, which would otherwise oxidize at the gaps or discontinuities lefi in the coating by previous casting methods.

Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

NASA Astrophysics Data System (ADS)

Myers, Astasia

2011-06-01

The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

Uranium from German Nuclear Power Projects of the 1940s--A Nuclear Forensic Investigation.

PubMed

Mayer, Klaus; Wallenius, Maria; Lützenkirchen, Klaus; Horta, Joan; Nicholl, Adrian; Rasmussen, Gert; van Belle, Pieter; Varga, Zsolt; Buda, Razvan; Erdmann, Nicole; Kratz, Jens-Volker; Trautmann, Norbert; Fifield, L Keith; Tims, Stephen G; Fröhlich, Michaela B; Steier, Peter

2015-11-02

Here we present a nuclear forensic study of uranium from German nuclear projects which used different geometries of metallic uranium fuel. Through measurement of the (230)Th/(234)U ratio, we could determine that the material had been produced in the period from 1940 to 1943. To determine the geographical origin of the uranium, the rare-earth-element content and the (87)Sr/(86)Sr ratio were measured. The results provide evidence that the uranium was mined in the Czech Republic. Trace amounts of (236)U and (239)Pu were detected at the level of their natural abundance, which indicates that the uranium fuel was not exposed to any major neutron fluence. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

DOEpatents

Roake, W.E.

1958-08-19

A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

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

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

FUEL ASSAY REACTOR

DOEpatents

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

DOE PAGES

Parfitt, David C.; Cooper, Michael William; Rushton, Michael J.D.; ...

2016-07-29

Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

OPTIMIZATION OF HETEROGENEOUS UTILIZATION OF THORIUM IN PWRS TO ENHANCE PROLIFERATION RESISTANCE AND REDUCE WASTE.

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

TODOSOW,M.; KAZIMI,M.

2004-08-01

Issues affecting the implementation, public perception and acceptance of nuclear power include: proliferation, radioactive waste, safety, and economics. The thorium cycle directly addresses the proliferation and waste issues, but optimization studies of core design and fuel management are needed to ensure that it fits within acceptable safety and economic margins. Typical pressurized water reactors, although loaded with uranium fuel, produce 225 to 275 kg of plutonium per gigawatt-year of operation. Although the spent fuel is highly radioactive, it nevertheless offers a potential proliferation pathway because the plutonium is relatively easy to separate, amounts to many critical masses, and does notmore » present any significant intrinsic barrier to weapon assembly. Uranium 233, on the other hand, produced by the irradiation of thorium, although it too can be used in weapons, may be ''denatured'' by the addition of natural, depleted or low enriched uranium. Furthermore, it appears that the chemical behavior of thoria or thoria-urania fuel makes it a more stable medium for the geological disposal of the spent fuel. It is therefore particularly well suited for a once-through fuel cycle. The use of thorium as a fertile material in nuclear fuel has been of interest since the dawn of nuclear power technology due to its abundance and to potential neutronic advantages. Early projects include homogeneous mixtures of thorium and uranium oxides in the BORAX-IV, Indian Point I, and Elk River reactors, as well as heterogeneous mixtures in the Shippingport seed-blanket reactor. However these projects were developed under considerably different circumstances than those which prevail at present. The earlier applications preceded the current proscription, for non-proliferation purposes, of the use of uranium enriched to more than 20 w/o in {sup 235}U, and has in practice generally prohibited the use of uranium highly enriched in {sup 235}U. They were designed when the expected burnup of light water fuel was on the order of 25 MWD/kgU--about half the present day value--and when it was expected that the spent fuel would be recycled to recover its fissile content.« less

The Gas Hills uranium district and some probable controls for ore deposition

USGS Publications Warehouse

Zeller, Howard Davis

1957-01-01

Uranium deposits occur in the upper coarse-grained facies of the Wind River formation of Eocene age in the Gas Hills district of the southern part of the Wind River Basin. Some of the principal deposits lie below the water table in the unoxidized zone and consist of uraninite and coffinite occurring as interstitial fillings in irregular blanket-like bodies. In the near-surface deposits that lie above the water table, the common yellow uranium minerals consist of uranium phosphates, silicates, and hydrous oxides. The black unoxidized uraninite -coffinite ores show enrichment of molybdenum, arsenic, and selenium when compared to the barren sandstone. Probable geologic controls for ore deposits include: 1) permeable sediments that allowed passage of ore-bearing solutions; 2) numerous faults that acted as impermeable barriers impounding the ore -bearing solutions; 3) locally abundant pyrite, carbonaceous material, and natuial gas containing hydrogen sulfide that might provide a favorable environment for precipitation of uranium. Field and laboratory evidence indicate that the uranium deposits in the Gas Hills district are very young and related to the post-Miocene to Pleistocene regional tilting to the south associated with the collapse of the Granite Mountains fault block. This may have stopped or reversed ground water movement from a northward (basinward) direction and alkaline ground water rich in carbonate could have carried the uranium into the favorable environment that induced precipitation.

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

NASA Astrophysics Data System (ADS)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

2015-12-01

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

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

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

An efficient molybdenum disulfide/cobalt diselenide hybrid catalyst for electrochemical hydrogen generation

PubMed Central

Gao, Min-Rui; Liang, Jin-Xia; Zheng, Ya-Rong; Xu, Yun-Fei; Jiang, Jun; Gao, Qiang; Li, Jun; Yu, Shu-Hong

2015-01-01

The electroreduction of water for sustainable hydrogen production is a critical component of several developing clean-energy technologies, such as water splitting and fuel cells. However, finding a cheap and efficient alternative catalyst to replace currently used platinum-based catalysts is still a prerequisite for the commercialization of these technologies. Here we report a robust and highly active catalyst for hydrogen evolution reaction that is constructed by in situ growth of molybdenum disulfide on the surface of cobalt diselenide. In acidic media, the molybdenum disulfide/cobalt diselenide catalyst exhibits fast hydrogen evolution kinetics with onset potential of −11 mV and Tafel slope of 36 mV per decade, which is the best among the non-noble metal hydrogen evolution catalysts and even approaches to the commercial platinum/carbon catalyst. The high hydrogen evolution activity of molybdenum disulfide/cobalt diselenide hybrid is likely due to the electrocatalytic synergistic effects between hydrogen evolution-active molybdenum disulfide and cobalt diselenide materials and the much increased catalytic sites. PMID:25585911

Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

1973-01-01

The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

NASA Astrophysics Data System (ADS)

Syarifah, Ratna Dewi; Suud, Zaki

2015-09-01

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

Early breakthrough of molybdenum and uranium in a permeable reactive barrier.

PubMed

Morrison, Stan J; Mushovic, Paul S; Niesen, Preston L

2006-03-15

A permeable reactive barrier (PRB) using zerovalent iron (ZVI) was installed at a site near Cañon City, CO, to treat molybdenum (Mo) and uranium (U) in groundwater. The PRB initially decreased Mo concentrations from about 4.8 to less than 0.1 mg/L; however, Mo concentrations in the ZVI increased to 2.0 mg/L after about 250 days and continued to increase until concentrations in the ZVI were about 4 times higherthan in the influent groundwater. Concentrations of U were reduced from 1.0 to less than 0.02 mg/L during the same period. Investigations of solid-phase samples indicate that (1) calcium carbonate, iron oxide, and sulfide minerals had precipitated in pores of the ZVI; (2) U and Mo were concentrated in the upgradient 5.1 cm of the ZVI; and (3) calcium was present throughout the ZVI accounting for up to 20.5% of the initial porosity. Results of a column test indicated that the ZVI from the PRB was still reactive for removing Mo and that removal rates were dependenton residence time and pH. The chemical evolution of the PRB is explained in four stages that present a progression from porous media flow through preferential flow and, finally, complete bypass of the ZVI.

The manufacture of LEU fuel elements at Dounreay

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

Gibson, J.

1997-08-01

Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

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

Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.

2012-07-31

Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed tomore » mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.« less

Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

NASA Technical Reports Server (NTRS)

Slaby, J. G.; Siegel, B. L.

1973-01-01

The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

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

J. Gan; D. Keiser; D. Wachs

2009-11-01

Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up tomore » ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.« less

Global Uranium And Thorium Resources: Are They Adequate To Satisfy Demand Over The Next Half Century?

NASA Astrophysics Data System (ADS)

Lambert, I. B.

2012-04-01

This presentation will consider the adequacy of global uranium and thorium resources to meet realistic nuclear power demand scenarios over the next half century. It is presented on behalf of, and based on evaluations by, the Uranium Group - a joint initiative of the OECD Nuclear Energy Agency and the International Atomic Energy Agency, of which the author is a Vice Chair. The Uranium Group produces a biennial report on Uranium Resources, Production and Demand based on information from some 40 countries involved in the nuclear fuel cycle, which also briefly reviews thorium resources. Uranium: In 2008, world production of uranium amounted to almost 44,000 tonnes (tU). This supplied approximately three-quarters of world reactor requirements (approx. 59,000 tU), the remainder being met by previously mined uranium (so-called secondary sources). Information on availability of secondary sources - which include uranium from excess inventories, dismantling nuclear warheads, tails and spent fuel reprocessing - is incomplete, but such sources are expected to decrease in market importance after 2013. In 2008, the total world Reasonably Assured plus Inferred Resources of uranium (recoverable at less than 130/kgU) amounted to 5.4 million tonnes. In addition, it is clear that there are vast amounts of uranium recoverable at higher costs in known deposits, plus many as yet undiscovered deposits. The Uranium Group has concluded that the uranium resource base is more than adequate to meet projected high-case requirements for nuclear power for at least half a century. This conclusion does not assume increasing replacement of uranium by fuels from reprocessing current reactor wastes, or by thorium, nor greater reactor efficiencies, which are likely to ameliorate future uranium demand. However, progressively increasing quantities of uranium will need to be mined, against a backdrop of the relatively small number of producing facilities around the world, geopolitical uncertainties and strong opposition to growth of nuclear power in a number of quarters - it is vital that the market provides incentives for exploration and development of environmentally sustainable mining operations. Thorium: World Reasonably Assured plus Inferred Resources of thorium are estimated at over 2.2 million tonnes, in hard rock and heavy mineral sand deposits. At least double this amount is considered to occur in as yet undiscovered thorium deposits. Currently, demand for thorium is insignificant, but even a major shift to thorium-fueled reactors would not make significant inroads into the huge resource base over the next half century.

Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

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

Bernard Jones; J. Blair Briggs; Leland Monteirth

A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well asmore » to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated uncertainty in the bias values were included in the overall uncertainty in benchmark keff values. Bias values were very small, ranging from 0.0004 ?k low to 0.0007 ?k low. Overall uncertainties range from ? 0.0018 to ? 0.0030. Major contributors to the overall uncertainty include uncertainty in the extrapolation to the uranium critical mass and the uranium density. Results are summarized in Figure 1. Figure 1. Experimental, Benchmark-Model, and MCNP/KENO Calculated Results The 32 configurations described and evaluated under ICSBEP Identifier HEU-MET-FAST-084 are judged to be acceptable for use as criticality safety benchmark experiments and should be valuable integral benchmarks for nuclear data testing of the various reflector materials. Details of the benchmark models, uncertainty analyses, and final results are given in this paper.« less

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

DART model for irradiation-induced swelling of uranium silicide dispersion fuel elements

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

Rest, J.; Hofman, G.L.

1999-04-01

Models for the interaction of uranium silicide dispersion fuels with an aluminum matrix, for the resultant reaction product swelling, and for the calculation of the stress gradient within the fuel particles are described within the context of DART fission-gas-induced swelling models. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by comparing DART calculations with irradiation data for the swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al in variously designed dispersion fuel elements.

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

Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

NASA Astrophysics Data System (ADS)

Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

2018-02-01

The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

NASA Technical Reports Server (NTRS)

Creagh, J. W. R.; Smith, J. R.

1973-01-01

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

NASA Technical Reports Server (NTRS)

Grisaffe, Salvatore J.; Caves, Robert M.

1964-01-01

An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

RECONDITIONING FUEL ELEMENTS

DOEpatents

Brandt, H.L.

1962-02-20

A process is given for decanning fuel elements that consist of a uranium core, an intermediate section either of bronze, silicon, Al-Si, and uranium silicide layers or of lead, Al-Si, and uranium silicide layers around said core, and an aluminum can bonded to said intermediate section. The aluminum can is dissolved in a solution of sodium hydroxide (9 to 20 wt%) and sodium nitrate (35 to 12 wt %), and the layers of the intermediate section are dissolved in a boiling sodium hydroxide solution of a minimum concentration of 50 wt%. (AEC) A method of selectively reducing plutonium oxides and the rare earth oxides but not uranium oxides is described which comprises placing the oxides in a molten solvent of zinc or cadmium and then adding metallic uranium as a reducing agent. (AEC)

WNA's worldwide overview on front-end nuclear fuel cycle growth and health, safety and environmental issues.

PubMed

Saint-Pierre, Sylvain; Kidd, Steve

2011-01-01

This paper presents the WNA's worldwide nuclear industry overview on the anticipated growth of the front-end nuclear fuel cycle from uranium mining to conversion and enrichment, and on the related key health, safety, and environmental (HSE) issues and challenges. It also puts an emphasis on uranium mining in new producing countries with insufficiently developed regulatory regimes that pose greater HSE concerns. It introduces the new WNA policy on uranium mining: Sustaining Global Best Practices in Uranium Mining and Processing-Principles for Managing Radiation, Health and Safety and the Environment, which is an outgrowth of an International Atomic Energy Agency (IAEA) cooperation project that closely involved industry and governmental experts in uranium mining from around the world. Copyright © 2010 Health Physics Society

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-07-07

Mining and Milling ................................................................................................7 Uranium Sales to India...carried out at Lucas Heights (see below). The nuclear fuel cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally... mining and milling stage. Commercial enrichment services are available in the United States, Europe, Russia, and Japan. Fuel fabrication services are

Atomic Fuel, Understanding the Atom Series. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is part of the "Understanding the Atom" series. Complete sets of the series are available free to teachers, schools, and public librarians who can make them available for reference or use by groups. Among the topics discussed are: What Atomic Fuel Is; The Odyssey of Uranium; Production of Uranium; Fabrication of Reactor…

Some Thermodynamic Features of Uranium-Plutonium Nitride Fuel in the Course of Burnup

NASA Astrophysics Data System (ADS)

Rusinkevich, A. A.; Ivanov, A. S.; Belov, G. V.; Skupov, M. V.

2017-12-01

Calculation studies on the effect of carbon and oxygen impurities on the chemical and phase compositions of nitride uranium-plutonium fuel in the course of burnup are performed using the IVTANTHERMO code. It is shown that the number of moles of UN decreases with increasing burnup level, whereas UN1.466, UN1.54, and UN1.73 exhibit a considerable increase. The presence of oxygen and carbon impurities causes an increase in the content of the UN1.466, UN1.54 and UN1.73 phases in the initial fuel by several orders of magnitude, in particular, at a relatively low temperature. At the same time, the presence of impurities abruptly reduces the content of free uranium in unburned fuel. Plutonium in the considered system is contained in form of Pu, PuC, PuC2, Pu2C3, and PuN. Plutonium carbides, as well as uranium carbides, are formed in small amounts. Most of the plutonium remains in the form of nitride PuN, whereas unbound Pu is present only in the areas with a low burnup level and high temperatures.

Comparison of Tungsten and Molybdenum Based Emitters for Advanced Thermionic Space Nuclear Reactors

NASA Astrophysics Data System (ADS)

Lee, Hsing H.; Dickinson, Jeffrey W.; Klein, Andrew C.; Lamp, Thomas R.

1994-07-01

Variations to the Advanced Thermionic Initiative thermionic fuel element are analyzed. Analysis included neutronic modeling with MCNP for criticality determination and thermal power distribution, and thermionic performance modeling with TFEHX. Changes to the original ATI configuration include the addition of W-HfC wire to the emitter for high temperature creep resistance improvement and substitution of molybdenum for the tungsten base material. Results from MCNP showed that all the tungsten used in the coating and base material must be 100% W-184 to obtain criticality. The presence of molybdenum in the emitter base affects the neutronic performance of the TFE by increasing the emitter neutron absorption cross section. Due to the reduced thermal conductivity for the molybdenum based emitter, a higher temperature is obtained resulting in a greater electrical power production. The thermal conductivity and resistivity of the composite emitter region were derived for the W-Mo composite and used in TFEHX.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

NASA Astrophysics Data System (ADS)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Empirical calibration of uranium releases in the terrestrial environment of nuclear fuel cycle facilities.

PubMed

Pourcelot, Laurent; Masson, Olivier; Saey, Lionel; Conil, Sébastien; Boulet, Béatrice; Cariou, Nicolas

2017-05-01

In the present paper the activity of uranium isotopes measured in plants and aerosols taken downwind of the releases of three nuclear fuel settlements was compared between them and with the activity measured at remote sites. An enhancement of 238 U activity as well as 235 U/ 238 U anomalies and 236 U are noticeable in wheat, grass, tree leaves and aerosols taken at the edge of nuclear fuel settlements, which show the influence of uranium chronic releases. Further plants taken at the edge of the studied sites and a few published data acquired in the same experimental conditions show that the 238 U activity in plants is influenced by the intensity of the U atmospheric releases. Assuming that 238 U in plant is proportional to the intensity of the releases, we proposed empirical relationships which allow to characterize the chronic releases on the ground. Other sources of U contamination in plants such as accidental releases and "delayed source" of uranium in soil are also discussed in the light of uranium isotopes signatures. Copyright © 2017 Elsevier Ltd. All rights reserved.

Separation of uranium from (Th,U)O.sub.2 solid solutions

DOEpatents

Chiotti, Premo; Jha, Mahesh Chandra

1976-09-28

Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2010 CFR

2010-04-01

...) Application of 50 percent test. The 50 percent test described in subparagraph (1) of this paragraph is applied... uranium concentrates (known in the industry as “yellow cake”), and nuclear fuel materials derived from the refining of uranium ore and uranium concentrates, or produced in a nuclear reaction, including— (a) Uranium...

26 CFR 1.993-3 - Definition of export property.

Code of Federal Regulations, 2011 CFR

2011-04-01

...) Application of 50 percent test. The 50 percent test described in subparagraph (1) of this paragraph is applied... uranium concentrates (known in the industry as “yellow cake”), and nuclear fuel materials derived from the refining of uranium ore and uranium concentrates, or produced in a nuclear reaction, including— (a) Uranium...

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

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

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

PRODUCTION OF PURIFIED URANIUM

DOEpatents

Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

1960-01-26

A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

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

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2009-08-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating “what if” scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level for U.S. nuclear power. The model is not intendedmore » as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., “reactor types” not individual reactors and “separation types” not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation of disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. We use Microsoft Excel 2003 and have not tested VISION with Microsoft Excel 2007. The VISION team uses both Powersim Studio 2005 and 2009 and it should work with either.« less

Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39)

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

PLYS, M.G.

2000-10-10

The purpose of this report is to provide a topical reference on the phenomena and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel (SNF) Project with specific applications to SNF Project processes and situations. Spent metallic uranium nuclear fuel is currently stored underwater at the K basins in the Hanford 100 area, and planned processing steps include: (1) At the basins, cleaning and placing fuel elements and scrap into stainless steel multi-canister overpacks (MCOs) holding about 6 MT of fuel apiece; (2) At nearby cold vacuum drying (CVD) stations, draining, vacuum drying, and mechanically sealing the MCOs; (3)more » Shipping the MCOs to the Canister Storage Building (CSB) on the 200 Area plateau; and (4) Welding shut and placing the MCOs for interim (40 year) dry storage in closed CSB storage tubes cooled by natural air circulation through the surrounding vault. Damaged fuel elements have exposed and corroded fuel surfaces, which can exothermically react with water vapor and oxygen during normal process steps and in off-normal situations, A key process safety concern is the rate of reaction of damaged fuel and the potential for self-sustaining or runaway reactions, also known as uranium fires or fuel ignition. Uranium metal and one of its corrosion products, uranium hydride, are potentially pyrophoric materials. Dangers of pyrophoricity of uranium and its hydride have long been known in the U.S. Department of Energy (Atomic Energy Commission/DOE) complex and will be discussed more below; it is sufficient here to note that there are numerous documented instances of uranium fires during normal operations. The motivation for this work is to place the safety of the present process in proper perspective given past operational experience. Steps in development of such a perspective are: (1) Description of underlying physical causes for runaway reactions, (2) Modeling physical processes to explain runaway reactions, (3) Validation of the method against experimental data, (4) Application of the method to plausibly explain operational experience, and (5) Application of the method to present process steps to demonstrate process safety and margin. Essentially, the logic above is used to demonstrate that runaway reactions cannot occur during normal SNF Project process steps, and to illustrate the depth of the technical basis for such a conclusion. Some off-normal conditions are identified here that could potentially lead to runaway reactions. However, this document is not intended to provide an exhaustive analysis of such cases. In summary, this report provides a ''toolkit'' of models and approaches for analysis of pyrophoricity safety issues at Hanford, and the technical basis for the recommended approaches. A summary of recommended methods appears in Section 9.0.« less

Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Apse, V. A.

2017-01-01

The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear power will allow us substantially to solve its problems, as well as to increase its export potential.

Geologic report on the San Rafael Swell Drilling Project, San Rafael Swell, Utah

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

Bluhm, C.T.; Rundle, J.G.

1981-08-01

Twenty-two holes totaling 34,874 feet (10,629.6 meters) were rotary and core drilled on the northern and western flanks of the San Rafael Swell to test fluvial-lacustrine sequences of the Morrison Formation and the lower part of the Chinle Formation. The objective of the project was to obtain subsurface data so that improved uranium resource estimates could be determined for the area. Although the Brushy Basin and the Salt Wash Members of the Morrison Formation are not considered favorable in this area for the occurrence of significant uranium deposits, uranium minerals were encountered in several of the holes. Some spotty ormore » very low-grade mineralization was also encountered in the White Star Trunk area. The lower part of the Chinle Formation is considered to be favorable for potentially significant uranium deposits along the west flank of the San Rafael Swell. One hole (SR-202) east of Ferron, Utah, intersected uranium, silver, molybdenum, and copper mineralization. More exploratory drilling in the vicinity of this hole is recommended. As a result of the study of many geochemical analyses and a careful determination of the lithology shown by drilling, a sabkha environment is suggested for the concentration of uranium, zinc, iron, lead, copper, silver, and perhaps other elements in parts of the Moody Canyon Member of the Moenkopi Formation.« less

Reconnaissance of uranium and copper deposits in parts of New Mexico, Colorado, Utah, Idaho, and Wyoming

USGS Publications Warehouse

Gott, Garland B.; Erickson, Ralph L.

1952-01-01

Because of the common association of uranium and copper in several of the commercial uranium deposits in the Colorado Plateau Province, a reconnaissance was made of several known deposits of copper disseminated through sandstone to determine whether they might be a source of uranium. In order to obtain more information regarding the relationship between copper, uranium and carbonaceous materials, some of the uraniferious asphaltrite deposits in the Shinarump conglomerate along the west flank of the San Rafael Swell were also investigated briefly. During this reconnaissance 18 deposits were examined in New Mexico, eight in Utah, two in Idaho, and one each in Wyoming and Colorado. No uranium deposits of commercial grade are associated with the copper deposits that were examined. The uraniferous asphaltites in the Shinarump conglomerate of Triassic age on the west flank of the San Rafael Swell, however, are promising from the standpoint of commercial uranium production. Spectrographic analyses of crude oil, asphalt, and bituminous shales show a rather consistent suite of trace metals including vanadium, nickel, copper, cobalt, chromium, lead zinc, and molybdenum. The similarity of the metal assemblage, including uranium of the San Rafael Swell asphaltites, to the metal assemblage in crude oil and other bituminous materials suggests that these metals were concentrated in the asphaltites from petroleum. However, the hypothesis that uranium minerals were already present before the hydrocarbons were introduced and that some sort of replacement or uranium minerals by carbon compounds was effected after the petroleum migrated into the uranium deposit should not be disregarded. The widespread association of uranium with asphaltic material suggests that it also may have been concentrated by some agency connected with the formation of petroleum. The problem of the association of uranium and other trace metals with hydrocarbons should be studied further both in the field and in the laboratory.

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

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

Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

2013-07-01

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in themore » Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)« less

Evaluation of Computed Tomography of Mock Uranium Fuel Rods at the Advanced Photon Source

DOE PAGES

Hunter, James F.; Brown, Donald William; Okuniewski, Maria

2015-06-01

This study discusses a multi-year effort to evaluate the utility of computed tomography at the Advanced Photon Source (APS) as a tool for non-destructive evaluation of uranium based fuel rods. The majority of the data presented is on mock material made with depleted uranium which mimics the x-ray attenuation characteristics of fuel rods while allowing for simpler handling. A range of data is presented including full thickness (5mm diameter) fuel rodlets, reduced thickness (1.8mm) sintering test samples, and pre/post irradiation samples (< 1mm thick). These data were taken on both a white beam (bending magnet) beamline and a high energy,more » monochromatic beamline. This data shows the utility of a synchrotron type source in the evealuation of manufacturing defects (pre-irradiation) and lays out the case for in situ CT of fuel pellet sintering. Finally, in addition data is shown from small post-irradiation samples and a case is made for post-irradiation CT of larger samples.« less

DPASV analytical technique for ppb level uranium analysis

NASA Astrophysics Data System (ADS)

Pal, Sangita; Singha, Mousumi; Meena, Sher Singh

2018-04-01

Determining uranium in ppb level is considered to be most crucial for reuse of water originated in nuclear industries at the time of decontamination of plant effluents generated during uranium (fuel) production, fuel rod fabrication, application in nuclear reactors and comparatively small amount of effluents obtained during laboratory research and developmental work. Higher level of uranium in percentage level can be analyzed through gravimetry, titration etc, whereas inductively coupled plasma-atomic energy spectroscopy (ICP-AES), fluorimeter are well suited for ppm level. For ppb level of uranium, inductively coupled plasma - mass spectroscopy (ICP-MS) or Differential Pulse Anodic Stripping Voltammetry (DPASV) serve the purpose. High precision, accuracy and sensitivity are the crucial for uranium analysis in trace (ppb) level, which are satisfied by ICP-MS and stripping voltammeter. Voltammeter has been found to be less expensive, requires low maintenance and is convenient for measuring uranium in presence of large number of other ions in the waste effluent. In this paper, necessity of uranium concentration quantification for recovery as well as safe disposal of plant effluent, working mechanism of voltammeter w.r.t. uranium analysis in ppb level with its standard deviation and a data comparison with ICP-MS has been represented.

Uranium Mines and Mills | RadTown USA | US EPA

EPA Pesticide Factsheets

2017-08-07

Uranium is used as nuclear fuel for electric power generation. U.S. mining industries can obtain uranium in two ways: mining or milling. Mining waste and mill tailings can contaminate water, soil and air if not disposed of properly.

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Layer, E.H. Jr.; Peet, C.S.

1962-01-23

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

Code of Federal Regulations, 2012 CFR

2012-01-01

... a discussion of the environmental significance of the data set forth in the table as weighed in the... 10 Energy 2 2012-01-01 2012-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the...

10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

Code of Federal Regulations, 2011 CFR

2011-01-01

... a discussion of the environmental significance of the data set forth in the table as weighed in the... 10 Energy 2 2011-01-01 2011-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the...

10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

Code of Federal Regulations, 2014 CFR

2014-01-01

... a discussion of the environmental significance of the data set forth in the table as weighed in the... 10 Energy 2 2014-01-01 2014-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the...

10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

Code of Federal Regulations, 2013 CFR

2013-01-01

... a discussion of the environmental significance of the data set forth in the table as weighed in the... 10 Energy 2 2013-01-01 2013-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the...

FUEL ELEMENT AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.

1961-04-25

A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the... Cycle Environmental Data, as the basis for evaluating the contribution of the environmental effects of...

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

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

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references.

On the equilibrium isotopic composition of the thorium-uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2016-12-01

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium-uranium-plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

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

Matthew Bunn; Steve Fetter; John P. Holdren

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recyclingmore » to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.« less

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

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

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spentmore » nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

Tags to Track Illicit Uranium and Plutonium

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

Haire, M. Jonathan; Forsberg, Charles W.

2007-07-01

With the expansion of nuclear power, it is essential to avoid nuclear materials from falling into the hands of rogue nations, terrorists, and other opportunists. This paper examines the idea of detection and attribution tags for nuclear materials. For a detection tag, it is proposed to add small amounts [about one part per billion (ppb)] of {sup 232}U to enriched uranium to brighten its radioactive signature. Enriched uranium would then be as detectable as plutonium and thus increase the likelihood of intercepting illicit enriched uranium. The use of rare earth oxide elements is proposed as a new type of 'attribution'more » tag for uranium and thorium from mills, uranium and plutonium fuels, and other nuclear materials. Rare earth oxides are chosen because they are chemically compatible with the fuel cycle, can survive high-temperature processing operations in fuel fabrication, and can be chosen to have minimal neutronic impact within the nuclear reactor core. The mixture of rare earths and/or rare earth isotopes provides a unique 'bar code' for each tag. If illicit nuclear materials are recovered, the attribution tag can identify the source and lot of nuclear material, and thus help police reduce the possible number of suspects in the diversion of nuclear materials based on who had access. (authors)« less

HIGH DENSITY NUCLEAR FUEL COMPOSITION

DOEpatents

Litton, F.B.

1962-07-17

ABS>A nuclear fuel consisting essentially of uranium monocarbide and containing 2.2 to 4.6 wt% carbon, 0.1 to 2.3 wt% oxygen, 0.05 to 2.5 wt% nitrogen, and the balance uranium was developed. The maximum oxygen content was less than one-half the carbon content by weight and the carbon, oxygen, and nitrogen are present as a single phase substituted solid solution of UC, C, O, and N. A method of preparing the fuel composition is described. (AEC)

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

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

Ruedig, Elizabeth; Johnson, Thomas E.

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA

DOE PAGES

Ruedig, Elizabeth; Johnson, Thomas E.

2015-08-30

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (asmore » nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6–8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42–0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y –1. As a result, higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear.« less

An evaluation of health risk to the public as a consequence of in situ uranium mining in Wyoming, USA.

PubMed

Ruedig, Elizabeth; Johnson, Thomas E

2015-12-01

In the United States there is considerable public concern regarding the health effects of in situ recovery uranium mining. These concerns focus principally on exposure to contaminants mobilized in groundwater by the mining process. However, the risk arising as a result of mining must be viewed in light of the presence of naturally occurring uranium ore and other constituents which comprise a latent hazard. The United States Environmental Protection Agency recently proposed new guidelines for successful restoration of an in situ uranium mine by limiting concentrations of thirteen groundwater constituents: arsenic, barium, cadmium, chromium, lead, mercury, selenium, silver, nitrate (as nitrogen), molybdenum, radium, total uranium, and gross α activity. We investigated the changes occurring to these constituents at an ISR uranium mine in Wyoming, USA by comparing groundwater quality at baseline measurement to that at stability (post-restoration) testing. Of the groundwater constituents considered, only uranium and radium-226 showed significant (p < 0.05) deviation from site-wide baseline conditions in matched-wells. Uranium concentrations increased by a factor of 5.6 (95% CI 3.6-8.9 times greater) while radium-226 decreased by a factor of about one half (95% CI 0.42-0.75 times less). Change in risk was calculated using the RESRAD (onsite) code for an individual exposed as a resident-farmer; total radiation dose to a resident farmer decreased from pre-to post-mining by about 5.2 mSv y(-1). Higher concentrations of uranium correspond to increased biomarkers of nephrotoxicity, however the clinical significance of this increase is unclear. Published by Elsevier Ltd.

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

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

Shmelev, A. N.; Kulikov, G. G., E-mail: ggkulikov@mephi.ru

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results aremore » analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction {sup 232+233+234}U and {sup 231}Pa are formulated. (1) The fuel cycle would shift from fissile {sup 235}U to {sup 233}U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of {sup 231}Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.« less

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.

2016-12-01

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U-Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction 232+233+234U and 231Pa are formulated. (1) The fuel cycle would shift from fissile 235U to 233U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most "protected" in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of 231Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.

The 1984 Goddard Space Flight Center Battery Workshop

NASA Technical Reports Server (NTRS)

Morrow, G. W. (Editor)

1985-01-01

Various topics related to spacecraft power supply systems are discussed. Regenerative fuel cells, lithium molybdenum batteries, nickel hydrogen batteries, nickel cadmium batteries, failure analysis, and performance testing are covered.

RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

DOEpatents

Gens, T.A.

1962-07-10

An improvement was made in a process of recovering uranium from a uranium-zirconium composition which was hydrochlorinated with gsseous hydrogen chloride at a temperature of from 350 to 800 deg C resulting in volatilization of the zirconium, as zirconium tetrachloride, and the formation of a uranium containing nitric acid insoluble residue. The improvement consists of reacting the nitric acid insoluble hydrochlorination residue with gaseous carbon tetrachloride at a temperature in the range 550 to 600 deg C, and thereafter recovering the resulting uranium chloride vapors. (AEC)

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 Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

Tuning Energetic Material Reactivity Using Surface Functionalization of Aluminum Fuels

DTIC Science & Technology

2012-10-30

analysis of three different thermites consisting of aluminum (Al) particles with and without surface functionalization combined with molybdenum...of thermites , aluminum synthesis, aluminum fluoropolymer combustion, acid coatings Keerti S. Kappagantula, Cory Farley, Michelle L. Pantoya, Jillian...Reactivity Using Surface Functionalization of Aluminum Fuels Report Title ABSTRACT Combustion analysis of three different thermites consisting of aluminum (Al

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

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

Miao, Yinbin; Ye, Bei; Mei, Zhigang

Uranium silicide (U 3Si 2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3Si 2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuelmore » material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3Si 2 fuel as an accident-tolerant alternative for uranium dioxide.« less

Data Validation Package December 2015 Groundwater Sampling at the Ambrosia Lake, New Mexico, Disposal Site March 2016

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

Tsosie, Bernadette; Johnson, Dick

The Long-Term Surveillance Plan for the Ambrosia Lake, New Mexico, Disposal Site does not require groundwater monitoring because groundwater in the uppermost aquifer is of limited use, and supplemental standards have been applied to the aquifer. However, at the request of the New Mexico Environment Department, the U.S. Department of Energy conducts annual monitoring at three locations: monitoring wells 0409, 0675, and 0678. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for US. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Monitoring Well 0409 was not sampled during this event because itmore » was dry. Water levels were measured at each sampled well. One duplicate sample was collected from location 0675. Groundwater samples from the two sampled wells were analyzed for the constituents listed in Table 1. Time-concentration graphs for selected analytes are included in this report. At well 0675, the duplicate results for total dissolved solids and for most metals (magnesium, molybdenum, potassium, selenium, sodium, and uranium) were outside acceptance criteria, which may indicate non-homogeneous conditions at this location. November 2014 results for molybdenum and uranium at well 0675 also were outside acceptance criteria. The well condition will be evaluated prior to the next sampling event.« less

Molecular dynamics simulation of thermal transport in UO 2 containing uranium, oxygen, and fission-product defects

DOE PAGES

Liu, Xiang -Yang; Cooper, Michael William D.; McClellan, Kenneth James; ...

2016-10-25

Uranium dioxide (UO 2) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defectmore » type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.« less

Molecular dynamics simulation of thermal transport in UO 2 containing uranium, oxygen, and fission-product defects

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

Liu, Xiang -Yang; Cooper, Michael William D.; McClellan, Kenneth James

Uranium dioxide (UO 2) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defectmore » type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.« less

Identification of Uranyl Minerals Using Oxygen K-Edge X Ray Absorption Spectroscopy

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

Ward, Jesse D.; Bowden, Mark E.; Resch, Charles T.

2016-03-01

Uranium analysis is consistently needed throughout the fuel cycle, from mining to fuel fabrication to environmental monitoring. Although most of the world’s uranium is immobilized as pitchblende or uraninite, there exists a plethora of secondary uranium minerals, nearly all of which contain the uranyl cation. Analysis of uranyl compounds can provide clues as to a sample’s facility of origin and chemical history. X-ray absorption spectroscopy is one technique that could enhance our ability to identify uranium minerals. Although there is limited chemical information to be gained from the uranium X-ray absorption edges, recent studies have successfully used ligand NEXAFS tomore » study the physical chemistry of various uranium compounds. This study extends the use of ligand NEXAFS to analyze a suite of uranium minerals. We find that major classes of uranyl compounds (carbonate, oxyhydroxide, silicate, and phosphate) exhibit characteristic lineshapes in the oxygen K-edge absorption spectra. As a result, this work establishes a library of reference spectra that can be used to classify unknown uranyl minerals.« less

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1964-03-24

A process of recovering plutonium from fuel by dissolution in molten KAlCl/sub 4/ double salt is described. Molten lithium chloride plus stannous chloride is added to reduce plutonium tetrachloride to the trichloride, which is dissolved in a lithium chloride phase while the uranium, as the tetrachloride, is dissolved in a double-salt phase. Separation of the two phases is discussed. (AEC)

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE PAGES

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

2015-06-21

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

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.

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

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

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Geology and ore deposits of the Section 23 Mine, Ambrosia Lake District, New Mexico

USGS Publications Warehouse

Granger, H.C.; Santos, E.S.

1982-01-01

The section 23 mine is one of about 18 large uranium mines opened in sandstones of the fluvial Westwater Canyon Member of the Jurassic Morrison Formation in the Ambrosia Lake mining district during the early 1960s. The Ambrosia Lake district is one of several mining districts within the Grants mineral belt, an elongate zone containing many uranium deposits along the southern flank of the San Juan basin. Two distinct types of ore occur in the mine. Primary ore occurs as peneconcordant layers of uranium-rich authigenic organic matter that impregnates parts of the reduced sandstone host rocks and which are typically elongate in an east-southeast direction subparallel both to the sedimentary trends and to the present-day regional strike of the strata. These are called prefault or trend ores because of their early genesis and their elongation and alinement. A second type of ore in the mine is referred to as postfault, stacked, or redistributed ore. Its genesis was similar to that of the roll-type deposits in Tertiary rocks of Wyoming and Texas. Oxidation, related to the development of a large tongue of oxidized rock extending from Gallup to Ambrosia Lake, destroyed much of the primary ore and redistributed it as massive accumulations of lower grade ores bordering the redox interface at the edge of the tongue. Host rocks in the southern half of sec. 23 (T. 14 N., R. 10 W.) are oxidized and contain only remnants of the original, tabular, organic-rich ore. Thick bodies of roll-type ore are distributed along the leading edge of the oxidized zone, and pristine primary ore is found only near the north edge of the section. Organic matter in the primary ore was derived from humic acids that precipitated in the pores of the sandstones and fixed uranium as both coffinite and urano-organic compounds. Vanadium, molybdenum, and selenium are also associated with the ore. The secondary or roll-type ores are essentially free of organic carbon and contain uranium both as coffinite and uraninite. They also contain vanadium and selenium but are virtually devoid of molybdenum. Although much has been learned about these deposits since the time this study was conducted, in 1966, a great deal more study will by required to completely elucidate their geologic history.

NEUTRONIC REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Finniston, H.M.; Plail, O.S.

1961-01-24

BS>A uranium body for use in a nuclear fission reactor is described. It has a homogeneous rod of uranium metal enclosed in an envelope of aluminum, wherein a thin metallic layer of higher melting point than aluminum and of relatively low competitive neutron absorption between the uranium and the aluminum is bonded to the uranium and to the aluminum of the sheath.

Phase discrimination of uranium oxides using laser-induced breakdown spectroscopy

NASA Astrophysics Data System (ADS)

Campbell, Keri R.; Wozniak, Nicholas R.; Colgan, James P.; Judge, Elizabeth J.; Barefield, James E.; Kilcrease, David P.; Wilkerson, Marianne P.; Czerwinski, Ken R.; Clegg, Samuel M.

2017-08-01

Nuclear forensics goals for characterizing samples of interest include qualitative and quantitative analysis of major and trace elements, isotopic analysis, phase identification, and physical analysis. These samples may include uranium oxides UO2, U3O8, and UO3, which play an important role in the front end of the nuclear fuel cycle, from mining to fuel fabrication. The focus of this study is to compare the ratios of the intensities of uranium and oxygen emission lines which can be used to distinguish between different uranium oxide materials using Laser-Induced Breakdown Spectroscopy (LIBS). Measurements at varying laser powers were made under an argon atmosphere at 585 Torr to ensure the oxygen emission intensity was originating from the sample, and not from the atmosphere. Fifteen uranium emission lines were used to compare experimental results with theoretical calculations in order to determine the plasma conditions. Using a laser energy of 26 mJ, the uranium lines 591.539 and 682.692 nm provide the highest degree of discrimination between the uranium oxides. The study presented here suggests that LIBS is useful for discriminating uranium oxide phases, UO2, U3O8, and UO3.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Timing of ore-related magmatism in the western Alaska Range, southwestern Alaska

USGS Publications Warehouse

Taylor, Ryan D.; Graham, Garth E.; Anderson, Eric D.; Selby, David

2014-01-01

This report presents isotopic age data from mineralized granitic plutons in an area of the Alaska Range located approximately 200 kilometers to the west-northwest of Anchorage in southwestern Alaska. Uranium-lead isotopic data and trace element concentrations of zircons were determined for 12 samples encompassing eight plutonic bodies ranging in age from approximately 76 to 57.4 millions of years ago (Ma). Additionally, a rhenium-osmium age of molybdenite from the Miss Molly molybdenum occurrence is reported (approx. 59 Ma). All of the granitic plutons in this study host gold-, copper-, and (or) molybdenum-rich prospects. These new ages modify previous interpretations regarding the age of magmatic activity and mineralization within the study area. The new ages show that the majority of the gold-quartz vein-hosting plutons examined in this study formed in the Late Cretaceous. Further work is necessary to establish the ages of ore-mineral deposition in these deposits.

TUNGSTEN BRONZE RELATED NON-NOBLE ELECTROCATALYSTS.

DTIC Science & Technology

FUEL CELLS, *CATALYSTS), (*OXYGEN, *ELECTRODES), (* SILICIDES , ELECTRODES), (*CARBIDES, ELECTRODES), (*TUNGSTEN COMPOUNDS, *ELECTROCHEMISTRY...CATALYSTS, TITANIUM COMPOUNDS, ZIRCONIUM COMPOUNDS, VANADIUM COMPOUNDS, NIOBIUM COMPOUNDS, TUNGSTEN COMPOUNDS, TANTALUM COMPOUNDS, MOLYBDENUM COMPOUNDS, SULFURIC ACID, CRYSTAL GROWTH, SODIUM COMPOUNDS

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

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

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Processing of irradiated, enriched uranium fuels at the Savannah River Plant

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

Hyder, M L; Perkins, W C; Thompson, M C

Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction withmore » dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.« less

Measures of the environmental footprint of the front end of the nuclear fuel cycle

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

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as wellmore » as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.« less

Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

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

Fishbone, L.G.; Moussalli, G.; Naegele, G.

1994-04-01

An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

Raman spectroscopic investigation of thorium dioxide-uranium dioxide (ThO₂-UO₂) fuel materials.

PubMed

Rao, Rekha; Bhagat, R K; Salke, Nilesh P; Kumar, Arun

2014-01-01

Raman spectroscopic investigations were carried out on proposed nuclear fuel thorium dioxide-uranium dioxide (ThO2-UO2) solid solutions and simulated fuels based on ThO2-UO2. Raman spectra of ThO2-UO2 solid solutions exhibited two-mode behavior in the entire composition range. Variations in mode frequencies and relative intensities of Raman modes enabled estimation of composition, defects, and oxygen stoichiometry in these compounds that are essential for their application. The present study shows that Raman spectroscopy is a simple, promising analytical tool for nondestructive characterization of this important class of nuclear fuel materials.

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

DOE PAGES

Davidson, Eva E.; Betzler, Benjamin R.; Chandler, David; ...

2017-08-01

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed andmore » benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.« less

77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... INFORMATION CONTACT: Glenn Tuttle, Office of Nuclear Material Safety and Safeguards, Division of Fuel Cycle...

RERTR-8 Irradiation Summary Report

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

D. M. Perez; M. A. Lillo; G. S. Chang

2011-12-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

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

Syarifah, Ratna Dewi, E-mail: syarifah.physics@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the additionmore » of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.« less

Status of the atomized uranium silicide fuel development at KAERI

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

Kim, C.K.; Kim, K.H.; Park, H.D.

1997-08-01

While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder.more » In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.« less

Heat pipe nuclear reactor for space power

NASA Technical Reports Server (NTRS)

Koening, D. R.

1976-01-01

A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

Epithermal uranium deposits in a volcanogenic context: the example of Nopal 1 deposit, Sierra de Pena Blanca, Mexico

NASA Astrophysics Data System (ADS)

Calas, G.; Angiboust, S.; Fayek, M.; Camacho, A.; Allard, T.; Agrinier, P.

2009-12-01

The Peña Blanca molybdenum-uranium field (Chihuahua, Mexico) exhibits over 100 airborne anomalies hosted in tertiary ignimbritic ash-flow tuffs (44 Ma) overlying the Pozos conglomerate and a sequence of Cretaceous carbonate rocks. Uranium occurrences are associated with breccia zones at the intersection of two or more fault systems. Periodic reactivation of these structures associated with Basin and Range and Rio Grande tectonic events resulted in the mobilization of U and other elements by meteoric fluids heated by geothermal activity. Trace element geochemistry (U, Th, REE) provides evidence for local mobilization of uranium under oxidizing conditions. In addition, O- and H-isotope geochemistry of kaolinite, smectite, opal and calcite suggests that argillic alteration proceeded at shallow depth with meteoric water at 25-75 °C. Focussed along breccia zones, fluids precipitated several generations of pyrite and uraninite together with kaolinite, as in the Nopal 1 mine, indicating that mineralization and hydrothermal alteration of volcanic tuffs are contemporaneous. Low δ34S values (~ -24.5 ‰) of pyrites intimately associated with uraninite suggest that the reducing conditions at the origin of the U-mineralization arise from biological activity. Later, the uplift of Sierra Pena Blanca resulted in oxidation and remobilization of uranium, as confirmed by the spatial distribution of radiation-induced defect centers in kaolinites. These data show that tectonism and biogenic reducing conditions can play a major role in the formation and remobilization of uranium in epithermal deposits. By comparison with the other uranium deposits at Sierra Pena Blanca and nearby Sierra de Gomez, Nopal 1 deposit is one of the few deposits having retained a reduced uranium mineralization.

Protected Nuclear Fuel Element

DOEpatents

Kittel, J. H.; Schumar, J. F.

1962-12-01

A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE PAGES

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

2017-05-08

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

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

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

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

NORTH ABSAROKA STUDY AREA, MONTANA.

USGS Publications Warehouse

Elliott, J.E.; Stotelmeyer, R.B.

1984-01-01

A mineral survey of the North Absaroka study area in Montana was conducted. The results of this survey indicate that parts of the area are extensively mineralized and that the area has potential for resources of gold, silver, copper, molybdenum, nickel, lead, zinc, platinum-group metals, uranium, iron, manganese, chromium, tungsten, and arsenic. Six areas of probable and substantiated mineral-resource potential were identified. The nature of the geologic terrain indicates that there is little likelihood for occurrence of oil, gas, coal, or geothermal resources.

Plasma Processing Systems for the Manufacture of Refractory Metals and their Alloys for Military Needs

DTIC Science & Technology

1978-10-09

melting point is around 4000*K. An exceedingly interesting feature of these solidification composites is the formation of fibrous MC type carbide ...the matrix could be refractory metal binary alloys with copper or uranium and the eutectic phase could be carbide of tungsten, * molybdenum, tantalum or...42 Accs -n or - *DTTI Tf Avn ! -7ll ’ i CrDi t , l’’*i,;. LIST OF FIGURES FIG. 1 Flow Diagram of Cemented Carbide Manufacture

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Catalysts and process for liquid hydrocarbon fuel production

DOEpatents

White, Mark G; Liu, Shetian

2014-12-09

The present invention provides a novel process and system in which a mixture of carbon monoxide and hydrogen synthesis gas, or syngas, is converted into hydrocarbon mixtures composed of high quality gasoline components, aromatic compounds, and lower molecular weight gaseous olefins in one reactor or step. The invention utilizes a novel molybdenum-zeolite catalyst in high pressure hydrogen for conversion, as well as a novel rhenium-zeolite catalyst in place of the molybdenum-zeolite catalyst, and provides for use of the novel catalysts in the process and system of the invention.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Helium interactions with alumina formed by atomic layer deposition show potential for mitigating problems with excess helium in spent nuclear fuel

NASA Astrophysics Data System (ADS)

Zhang, Shenli; Yu, Erick; Gates, Sean; Cassata, William S.; Makel, James; Thron, Andrew M.; Bartel, Christopher; Weimer, Alan W.; Faller, Roland; Stroeve, Pieter; Tringe, Joseph W.

2018-02-01

Helium gas accumulation from alpha decay during extended storage of spent fuel has potential to compromise the structural integrity the fuel. Here we report results obtained with surrogate nickel particles which suggest that alumina formed by atomic layer deposition can serve as a low volume-fraction, uniformly-distributed phase for retention of helium generated in fuel particles such as uranium oxide. Thin alumina layers may also form transport paths for helium in the fuel rod, which would otherwise be impermeable. Micron-scale nickel particles, representative of uranium oxide particles in their low helium solubility and compatibility with the alumina synthesis process, were homogeneously coated with alumina approximately 3-20 nm by particle atomic layer deposition (ALD) using a fluidized bed reactor. Particles were then loaded with helium at 800 °C in a tube furnace. Subsequent helium spectroscopy measurements showed that the alumina phase, or more likely a related nickel/alumina interface structure, retains helium at a density of at least 1017 atoms/cm3. High resolution transmission electron microscopy revealed that the thermal treatment increased the alumina thickness and generated additional porosity. Results from Monte Carlo simulations on amorphous alumina predict the helium retention concentration at room temperature could reach 1021 atoms/cm3 at 400 MPa, a pressure predicted by others to be developed in uranium oxide without an alumina secondary phase. This concentration is sufficient to eliminate bubble formation in the nuclear fuel for long-term storage scenarios, for example. Measurements by others of the diffusion coefficient in polycrystalline alumina indicate values several orders of magnitude higher than in uranium oxide, which then can also allow for helium transport out of the spent fuel.

Reconnaissance for radioactive deposits in eastern Alaska, 1952

USGS Publications Warehouse

Nelson, Arthur Edward; West, Walter S.; Matzko, John J.

1954-01-01

Reconnaissance for radioactive deposits was conducted in selected areas of eastern Alaska during 1952. Examination of copper, silver, and molybdenum occurrences and of a reported nickel prospect in the Slana-Nabesna and Chisana districts in the eastern Alaska Range revealed a maximum radioactivity of about 0.003 percent equivalent uranium. No appreciable radioactivity anomolies were indicated by aerial and foot traverses in the area. Reconnaissance for possible lode concentrations of uranium minerals in the vicinity of reported fluoride occurrences in the Hope Creek and Miller House-Circle Hot Springs areas of the Circle quadrangle and in the Fortymile district found a maximum of 0.055 percent equivalent uranium in a float fragment of ferruginous breccia in the Hope Creek area; analysis of samples obtained in the vicinity of the other fluoride occurrences showed a maximum of only 0.005 percent equivalent uranium. No uraniferous loads were discovered in the Koyukuk-Chandalar region, nor was the source of the monazite, previously reported in the placer concentrates from the Chandalar mining district, located. The source of the uranotheorianite in the placers at Gold Bench on the South Fork of the Koyukuk River was not found during a brief reconaissance, but a placer concentrate was obtained that contains 0.18 percent equivalent uranium. This concentrate is about ten times more radioactive than concentrates previously available from the area.

Preliminary investigation of the elemental variation and diagenesis of a tabular uranium deposit, La Sal Mine, San Juan County, Utah

USGS Publications Warehouse

Brooks, Robert A.; Campbell, John A.

1976-01-01

Ore in the La Sal mine, San Juan County, Utah, occurs as a typical tabular-type uranium deposit of the-Colorado Plateau. Uranium-vanadium occurs in the Salt Wash Member of the Jurassic Morrison Formation. Chemical and petrographic analyses were used to determine elemental variation and diagenetic aspects across the orebody. Vanadium is concentrated in the dark clay matrix, which constitutes visible ore. Uranium content is greater above the vanadium zone. Calcium, carbonate carbon, and lead show greater than fifty-fold increase across the ore zone, whereas copper and organic carbon show only a several-fold increase. Large molybdenum concentrations are present in and above the tabular layer, and large selenium concentrations occur below the uranium zone within the richest vanadium zone. Iron is enriched in the vanadium horizon. Chromium is depleted from above the ore and strongly enriched below. Elements that vary directly with the vanadium content include magnesium, iron, selenium, zirconium, strontium, titanium, lead, boron, yttrium, and scandium. The diagenetic sequence is as follows: (1) formation of secondary quartz overgrowths as cement; (2) infilling and lining of remaining pores with amber opaline material; (3) formation of vanadium-rich clay matrix, which has replaced overgrowths as well as quartz grains; (4) replacement of overgrowths and detrital grains by calcite; (5) infilling of pores with barite and the introduction of pyrite and marcasite.

The effect of thermomechanical processing on second phase particle redistribution in U-10 wt%Mo

NASA Astrophysics Data System (ADS)

Hu, Xiaohua; Wang, Xiaowo; Joshi, Vineet V.; Lavender, Curt A.

2018-03-01

The multi-pass hot-rolling process of an annealed uranium-10 wt% molybdenum (U10Mo) coupon was studied by plane-strain compression finite element modeling. As-cast U10Mo typically contains second phase particles such as uranium carbides (UC) and silicides along the grain boundaries. The volume fraction of UC is typically large, while the other phases can be redissolved in the matrix by certain heat treatments. The UC particle distribution is important due to its influence on the recrystallization processes (particle stimulated nucleation) that occur during annealing between rolling passes. Unfavorable particle distribution and fracture after rolling can affect the grain size and also influence the fuel performance in the reactor. A statistical method, i.e., the two-point correlation function (2PCF), was used to analyze the carbide particle distribution after each rolling reduction. The hot rolling simulation results show that the alignment of UC particles along grain boundaries will rotate during rolling until it is parallel to the rolling direction, to form stringer-like distributions which are typically observed in rolled products that contain inclusions. 2PCF analysis shows that the interparticle spacing shrinks along the normal direction (ND) and increases along the rolling direction (RD). The simulated particle distribution is very similar to that measured experimentally for similar rolling reductions. The magnitudes of major peaks of 2PCF along the ND decrease after large reduction. The locations of major peaks indicate the inter-stringer distances. Many more small peaks appear for the 2PCF along the RD, and this is related to the neighboring particles within stringers, which are along the RD.

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

NASA Astrophysics Data System (ADS)

Hallman, Luther, Jr.

Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

JPRS Report, Science & Technology, Japan

DTIC Science & Technology

1987-11-12

Change (4) Future Direction Anyway, it has become almost clear that the effect of power recovery cannot be expected from the insulation of...process spent fuels in greater safety and to recover the uranium or plutonium from spent fuels for effective reapplication. In 1974, the PNC began...constructed to serve as a pilot plant that could be used to establish reprocessing technology for the next practical stage. 32 As for enriched uranium

Partially Ionized Plasmas, Including the Third Symposium on Uranium Plasmas

NASA Technical Reports Server (NTRS)

Krishnan, M.

1976-01-01

Fundamentals of both electrically and fission generated plasmas are discussed. Research in gaseous fuel reactors using uranium hexafluoride is described and other partially ionized plasma applications are discussed.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Mixed uranium dicarbide and uranium dioxide microspheres and process of making same

DOEpatents

Stinton, David P.

1983-01-01

Nuclear fuel microspheres are made by sintering microspheres containing uranium dioxide and uncombined carbon in a 1 mole percent carbon monoxide/99 mole percent argon atmosphere at 1550.degree. C. and then sintering the microspheres in a 3 mole percent carbon monoxide/97 mole percent argon atmosphere at the same temperature.

Carbothermic synthesis of 820 μm uranium nitride kernels: Literature review, thermodynamics, analysis, and related experiments

NASA Astrophysics Data System (ADS)

Lindemer, T. B.; Voit, S. L.; Silva, C. M.; Besmann, T. M.; Hunt, R. D.

2014-05-01

The US Department of Energy is developing a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with uranium nitride (UN) kernels with diameters near 825 μm. This effort explores factors involved in the conversion of uranium oxide-carbon microspheres into UN kernels. An analysis of previous studies with sufficient experimental details is provided. Thermodynamic calculations were made to predict pressures of carbon monoxide and other relevant gases for several reactions that can be involved in the conversion of uranium oxides and carbides into UN. Uranium oxide-carbon microspheres were heated in a microbalance with an attached mass spectrometer to determine details of calcining and carbothermic conversion in argon, nitrogen, and vacuum. A model was derived from experiments on the vacuum conversion to uranium oxide-carbide kernels. UN-containing kernels were fabricated using this vacuum conversion as part of the overall process. Carbonitride kernels of ∼89% of theoretical density were produced along with several observations concerning the different stages of the process.

Preparation, certification and validation of a stable solid spike of uranium and plutonium coated with a cellulose derivative for the measurement of uranium and plutonium content in dissolved nuclear fuel by isotope dilution mass spectrometry.

PubMed

Surugaya, Naoki; Hiyama, Toshiaki; Verbruggen, André; Wellum, Roger

2008-02-01

A stable solid spike for the measurement of uranium and plutonium content in nitric acid solutions of spent nuclear fuel by isotope dilution mass spectrometry has been prepared at the European Commission Institute for Reference Materials and Measurements in Belgium. The spike contains about 50 mg of uranium with a 19.838% (235)U enrichment and 2 mg of plutonium with a 97.766% (239)Pu abundance in each individual ampoule. The dried materials were covered with a thin film of cellulose acetate butyrate as a protective organic stabilizer to resist shocks encountered during transportation and to eliminate flaking-off during long-term storage. It was found that the cellulose acetate butyrate has good characteristics, maintaining a thin film for a long time, but readily dissolving on heating with nitric acid solution. The solid spike containing cellulose acetate butyrate was certified as a reference material with certified quantities: (235)U and (239)Pu amounts and uranium and plutonium amount ratios, and was validated by analyzing spent fuel dissolver solutions of the Tokai reprocessing plant in Japan. This paper describes the preparation, certification and validation of the solid spike coated with a cellulose derivative.

Oxygen potential of uranium--plutonium oxide as determined by controlled- atmosphere thermogravimetry

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

Swanson, Gerald C.

1975-10-01

The oxygen-to-metal atom ratio, or O/M, of solid solution uranium- plutonium oxide reactor fuel is a measure of the concentration of crystal defects in the oxide which affect many fuel properties, particularly, fuel oxygen potential. Fabrication of a high-temperature oxygen electrode, employing an electro-active tip of oxygen-deficient solid-state electrolyte, intended to confirm gaseous oxygen potentials is described. Uranium oxide and plutonium oxide O/M reference materials were prepared by in situ oxidation of high purity metals in the thermobalance. A solid solution uranium-plutonium oxide O/M reference material was prepared by alloying the uranium and plutonium metals in a yttrium oxide cruciblemore » at 1200°C and oxidizing with moist He at 250°C. The individual and solid solution oxides were isothermally equilibrated with controlled oxygen potentials between 800 and 1300°C and the equilibrated O/ M ratios calculated with corrections for impurities and buoyancy effects. Use of a reference oxygen potential of -100 kcal/mol to produce an O/M of 2.000 is confirmed by these results. However, because of the lengthy equilibration times required for all oxides, use of the O/M reference materials rather than a reference oxygen potential is recommended for O/M analysis methods calibrations.« less

Some neutron and gamma radiation characteristics of plutonium cermet fuel for isotopic power sources

NASA Technical Reports Server (NTRS)

Neff, R. A.; Anderson, M. E.; Campbell, A. R.; Haas, F. X.

1972-01-01

Gamma and neutron measurements on various types of plutonium sources are presented in order to show the effects of O-17, O-18 F-19, Pu-236, age of the fuel, and size of the source on the gamma and neutron spectra. Analysis of the radiation measurements shows that fluorine is the main contributor to the neutron yields from present plutonium-molybdenum cermet fuel, while both fluorine and Pu-236 daughters contribute significantly to the gamma ray intensities.

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

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

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

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

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

Lundberg, L. B.; Hobbins, R. R.

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

Looking North at Uranium recovery Recycle Tanks in Red Room ...

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

Looking North at Uranium recovery Recycle Tanks in Red Room in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

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

B.R. Westphal; J.C. Price; R.D. Mariani

The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results andmore » conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.« less

Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

2018-06-01

Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

Safety and Regulatory Issues of the Thorium Fuel Cycle

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

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

Potassium-Promoted Molybdenum Carbide as a Highly Active and Selective Catalyst for CO2 Conversion to CO.

PubMed

Porosoff, Marc D; Baldwin, Jeffrey W; Peng, Xi; Mpourmpakis, Giannis; Willauer, Heather D

2017-06-09

The high concentration of CO 2 bound in seawater represents a significant opportunity to extract and use this CO 2 as a C 1 feedstock for synthetic fuels. Using an existing process, CO 2 and H 2 can be concurrently extracted from seawater and then catalytically reacted to produce synthetic fuel. Hydrogenating CO 2 directly into liquid hydrocarbons is exceptionally difficult, but by first identifying a catalyst for selective CO production through the reverse water-gas shift (RWGS) reaction, CO can then be hydrogenated to fuel through Fischer-Tropsch (FT) synthesis. Results of this study demonstrate that potassium-promoted molybdenum carbide supported on γ-Al 2 O 3 (K-Mo 2 C/γ-Al 2 O 3 ) is a low-cost, stable, and highly selective catalyst for RWGS over a wide range of conversions. These findings are supported by X-ray diffraction, scanning electron microscopy with energy-dispersive X-ray spectroscopy, X-ray photoelectron spectroscopy, and density functional theory calculations. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Sustained Recycle in Light Water and Sodium-Cooled Reactors

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

Steven J. Piet; Samuel E. Bays; Michael A. Pope

2010-11-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-03-01

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Returnmore » Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.« less

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

Lombardo, N.J.; Marseille, T.J.; White, M.D.

TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic inmore » form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.« less

Looking Northwest at Uranium Dryers Along North Side of Green ...

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

Looking Northwest at Uranium Dryers Along North Side of Green Room in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

Design of a Fission 99 Mo Recovery Process and Implications toward Mo Adsorption Mechanism on Titania and Alumina Sorbents

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

Stepinski, Dominique C.; Youker, Amanda J.; Krahn, Elizabeth O.

2017-03-01

Molybdenum-99 is a parent of the most widely used medical isotope technetium-99m. Proliferation concerns have prompted development of alternative Mo production methods utilizing low enriched uranium. Alumina and titania sorbents were evaluated for separation of Mo from concentrated uranyl nitrate solutions. System, mass transfer, and isotherm parameters were determined to enable design of Mo separation processes under a wide range of conditions. A model-based approach was utilized to design representative commercial-scale column processes. The designs and parameters were verified with bench-scale experiments. The results are essential for design of Mo separation processes from irradiated uranium solutions, selection of support materialmore » and process optimization. Mo uptake studies show that adsorption decreases with increasing concentration of uranyl nitrate; howeveL, examination of Mo adsorption as a function of nitrate ion concentration shows no dependency, indicating that uranium competes with Mo for adsorption sites. These results are consistent with reports indicating that Mo forms inner-sphere complexes with titania and alumina surface groups.« less

Direct measurement of 235U in spent fuel rods with Gamma-ray mirrors

NASA Astrophysics Data System (ADS)

Ruz, J.; Brejnholt, N. F.; Alameda, J. B.; Decker, T. A.; Descalle, M. A.; Fernandez-Perea, M.; Hill, R. M.; Kisner, R. A.; Melin, A. M.; Patton, B. W.; Soufli, R.; Ziock, K.; Pivovaroff, M. J.

2015-03-01

Direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable non-destructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect Kα and Kβ fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

METHOD OF MAKING WIRE FUEL ELEMENTS

DOEpatents

Zambrow, J.L.

1960-08-01

A method is given for making a nuclear reactor fuel element in the form of a uranium-bearing wire clad with zirconium. A uranium bar is enclosed in a zirconium sheath which is coated with an oxide of magnesium, beryllium, or zirconium. The sheathed bar is then placed in a steel tube and reduced to the desired diameter by swaging at 800 to 900 deg C, after which the steel and oxide are removed.

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

NASA Astrophysics Data System (ADS)

Ulrich, J. C.; Guilhen, S. N.; Cotrim, M. E. B.; Pires, M. A. F.

2018-03-01

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

Plutonium-uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

NASA Astrophysics Data System (ADS)

Degueldre, C.; Martin, M.; Kuri, G.; Grolimund, D.; Borca, C.

2011-09-01

Plutonium-uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg -1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (<5%) Pu(V) or Pu(VI) can be detected while the fuel could undergo slight oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

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

Rice, Francine Joyce; Stempien, John Dennis

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within amore » specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.« less

Molybdenum nitrides as oxygen reduction reaction catalysts: Structural and electrochemical studies

DOE PAGES

Cao, Bingfei; Neuefeind, Joerg C.; Adzic, Radoslav R.; ...

2015-02-09

Monometallic (δ-MoN, Mo 5N 6, and Mo 2N) and bimetallic molybdenum nitrides (Co 0.6Mo 1.4N 2) were investigated as electrocatalysts for the oxygen reduction reaction (ORR), which is a key half-reaction in hydrogen fuel cells. Monometallic hexagonal molybdenum nitrides are found to exhibit improved activities over rock salt type molybdenum nitride (γ-Mo 2N), suggesting that improvements are due to either the higher molybdenum valence or a more favorable coordination environment in the hexagonal structures. Further enhancements in activity were found for hexagonal bimetallic cobalt molybdenum nitride (Co 0.6Mo 1.4N 2), resulting in a modest onset potential of 0.713 V versusmore » reversible hydrogen electrode (RHE). Co 0.6Mo 1.4N 2 exhibits good stability in acidic environments, and in the potential range lower than 0.5 V versus RHE, the ORR appears to proceed via a four-electron mechanism based on the analysis of rotating disc electrode results. A redetermination of the structures of the binary molybdenum nitrides was carried out using neutron diffraction data, which is far more sensitive to nitrogen site positions than X-ray diffraction data. In conclusion, the revised monometallic hexagonal nitride structures all share many common features with the Co 0.6Mo 1.4N 2 structure, which has alternating layers of cations in octahedral and trigonal prismatic coordination, and are thus not limited to only trigonal prismatic Mo environments (as was originally postulated for δ-MoN).« less

PLUTONIUM-URANIUM-TITANIUM ALLOYS

DOEpatents

Coffinberry, A.S.

1959-07-28

A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

40 CFR 190.12 - Effective date.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Uranium Fuel Cycle § 190.12 Effective date. (a) The standards in § 190.10(a) shall be effective December 1, 1979, except that for doses arising from operations associated with the milling of uranium ore the...

Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

NASA Astrophysics Data System (ADS)

Grande, Lisa Christine

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

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

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

Radionuclides and trace metals in Canadian moose near uranium mines: comparison of radiation doses and food chain transfer with cattle and caribou.

PubMed

Thomas, Patricia; Irvine, James; Lyster, Jane; Beaulieu, Rhys

2005-05-01

Tissues from 45 moose and 4 cattle were collected to assess the health of country foods near uranium mines in northern Saskatchewan. Bone, liver, kidney, muscle and rumen contents were analyzed for uranium, radium-226 (226Ra), lead-210 (210Pb), and polonium-210 (210Po). Cesium-137 (137Cs), potassium-40 (40K), and 27 trace metals were also measured in some tissues. Within the most active mining area, Po in liver and muscle declined significantly with distance from tailings, possibly influenced by nearby natural uranium outcrops. Moose from this area had significantly higher 226Ra, 210Pb, 210Po, and 137Cs in some edible soft tissues vs. one control area. However, soil type and diet may influence concentrations as much as uranium mining activities, given that a) liver levels of uranium, 226Ra, and 210Po were similar to a second positive control area with mineral-rich shale hills and b) 210Po was higher in cattle kidneys than in all moose. Enhanced food chain transfer from rumen contents to liver was found for selenium in the main mining area and for copper, molybdenum and cadmium in moose vs. cattle. Although radiological doses to moose in the main mining area were 2.6 times higher than doses to control moose or cattle, low moose intakes yielded low human doses (0.0068 mSv y(-1)), a mere 0.3% of the dose from intake of caribou (2.4 mSv y(-1)), the dietary staple in the area.

Comparison of actinide production in traveling wave and pressurized water reactors

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

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

New Fiber Materials with Sorption Capacity at 5.0 g-U/kg Adsorbent under Marine Testing Conditions

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

Saito, Tomonori; Brown, S.; Das, Sadananda

The Fuel Resources program of the Fuel Cycle Research and Development program of the Office of Nuclear Energy (NE) has focused on assuring that nuclear fuel resources are available in the United States for a long term. An immense source of uranium is seawater, which contains an estimated amount of 4.5 billion tonnes of dissolved uranium. Extraction of the uranium resource in seawater can provide a price cap and ensure centuries of uranium supply for future nuclear energy production. NE initiated a multidisciplinary program with participants from national laboratories, universities, and research institutes to enable technical breakthroughs related to uraniummore » recovery from seawater. The goal is to develop advanced adsorbents to make the seawater uranium recovery technology a cost competitive, viable technology. Under this program, Oak Ridge National Laboratory (ORNL) has developed several novel adsorbents, which enhanced the uranium capacity 4-5 times from the state-of-the art Japanese adsorbents. Uranium exists uniformly at a concentration of ~3.3 ppb in seawater. Because of the vast volume of the oceans, the total estimated amount of uranium in seawater is approximately 1000 times larger than its amount in terrestrial resources. However, due to the low concentration, a significant challenge remains for making the extraction of uranium from seawater a commercially viable alternative technology. The biggest challenge for this technology to overcome to efficiently reduce the extraction cost is to develop adsorbents with increased uranium adsorption capacity. Two major approaches were investigated for synthesizing novel adsorbents with enhanced uranium adsorption capacity. One method utilized conventional radiation induced graft polymerization (RIGP) to synthesize adsorbents on high-surface area trunk fibers and the other method utilized a chemical grafting technique, atom-transfer radical polymerization (ATRP). Both approaches have shown promising uranium extraction capacities: RIGP adsorbent achieved 5.00 ± 0.15 g U/kg-ads., while ATRP adsorbent achieved 6.56 ± 0.33 g U/kg-ads., after 56 days of seawater exposure. These achieved values are the highest adsorption capacities ever reported for uranium extraction from seawater. The study successfully demonstrated new fiber materials with sorption capacity at 5.0 g-U/kg adsorbent under marine testing conditions. Further optimization, investigation of other new materials as well as deepening our understanding will develop adsorbents that have even higher uranium adsorption capacity, increased selectivity, and faster kinetics.« less

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

2013-06-01

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O2 lattice in an irradiated (60 MW d kg-1) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (˜0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am3+ species within an [AmO8]13- coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix.

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

Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael

Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K 2(UO 2) 3O 4·4H 2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarilymore » UO 3·2H 2O, with minor phases of U 3O 8 and UO 2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

Production of nanocrystalline metal powders via combustion reaction synthesis

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

Frye, John G.; Weil, Kenneth Scott; Lavender, Curt A.

Nanocrystalline metal powders comprising tungsten, molybdenum, rhenium and/or niobium can be synthesized using a combustion reaction. Methods for synthesizing the nanocrystalline metal powders are characterized by forming a combustion synthesis solution by dissolving in water an oxidizer, a fuel, and a base-soluble, ammonium precursor of tungsten, molybdenum, rhenium, or niobium in amounts that yield a stoichiometric burn when combusted. The combustion synthesis solution is then heated to a temperature sufficient to substantially remove water and to initiate a self-sustaining combustion reaction. The resulting powder can be subsequently reduced to metal form by heating in a reducing gas environment.

A top-down assessment of energy, water and land use in uranium mining, milling, and refining

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

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Land, water and energy use are key measures of the sustainability of uranium production into the future. As the most attractive, accessible deposits are mined out, future discoveries may prove to be significantly, perhaps unsustainably, more intensive consumers of environmental resources. A number of previous attempts have been made to provide empirical relationships connecting these environmental impact metrics to process variables such as stripping ratio and ore grade. These earlier attempts were often constrained by a lack of real world data and perform poorly when compared against data from modern operations. This paper conditions new empirical models of energy, watermore » and land use in uranium mining, milling, and refining on contemporary data reported by operating mines. It shows that, at present, direct energy use from uranium production represents less than 1% of the electrical energy produced by the once-through fuel cycle. Projections of future energy intensity from uranium production are also possible by coupling the empirical models with estimates of uranium crustal abundance, characteristics of new discoveries, and demand. The projections show that even for the most pessimistic of scenarios considered, by 2100, the direct energy use from uranium production represents less than 3% of the electrical energy produced by the contemporary once-through fuel cycle.« less

Growth of the interaction layer around fuel particles in dispersion fuel

NASA Astrophysics Data System (ADS)

Olander, D.

2009-01-01

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

Thermionic System Evaluation Test: Ya-21U System Topaz International Program

DTIC Science & Technology

1996-07-01

by enriched uranium dioxide (U02) fuel pellets, as illustrated by Figure 5. The work section of the converter contained 34 TFEs that provided power...power system. This feature permitted transportation of the highly enriched uranium oxide fuel in separate containers from the space power system and...by Figure 8. The radial reflector contained three safety and nine control drums. Each drum contained a section of boron carbide (B4C) neutron poison

Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

DOE PAGES

Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; ...

2015-04-13

Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K 2(UO 2) 3O 4·4H 2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarilymore » UO 3·2H 2O, with minor phases of U 3O 8 and UO 2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

DOEpatents

Handwerk, J.H.; Noland, R.A.; Walker, D.E.

1957-09-10

In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, Eddie C.; Miller, William E.; Laidler, James J.

1997-01-01

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two.

Electrorefining cell with parallel electrode/concentric cylinder cathode

DOEpatents

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1997-07-22

A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt. Scrapers located on each anode basket abrade and remove the spent fuel deposits on the surfaces of the inner and outer cathode cylinders, with the spent fuel falling to the bottom of the cell for removal. Cell resistance is reduced and uranium deposition rate enhanced by increasing the electrode area and reducing the anode-cathode spacing. Collection efficiency is enhanced by trapping and recovery of uranium dendrites scrapped off of the cylindrical cathodes which may be greater in number than two. 12 figs.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

1992-08-25

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

Uranium chloride extraction of transuranium elements from LWR fuel

DOEpatents

Miller, William E.; Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

METAL SHEATHED BODIES

DOEpatents

Finniston, H.M.; Wyatt, L.M.; Plail, O.S.

1961-06-27

An aluminum-cased uranium fuel element is patented for use in nuclear reactors. A layer of a substance such as graphite or a metallic film, preferably of relatively low thermal-neutron capture cross section, between the uranium and aluminum prevents their interdiffusion.

Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

ERIC Educational Resources Information Center

Settle, Frank A.

2009-01-01

The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

Carbothermic Synthesis of 820 m UN Kernels: Literature Review, Thermodynamics, Analysis, and Related Experiments

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

Lindemer, Terrence; Voit, Stewart L; Silva, Chinthaka M

2014-01-01

The U.S. Department of Energy is considering a new nuclear fuel that would be less susceptible to ruptures during a loss-of-coolant accident. The fuel would consist of tristructural isotropic coated particles with large, dense uranium nitride (UN) kernels. This effort explores many factors involved in using gel-derived uranium oxide-carbon microspheres to make large UN kernels. Analysis of recent studies with sufficient experimental details is provided. Extensive thermodynamic calculations are used to predict carbon monoxide and other pressures for several different reactions that may be involved in conversion of uranium oxides and carbides to UN. Experimentally, the method for making themore » gel-derived microspheres is described. These were used in a microbalance with an attached mass spectrometer to determine details of carbothermic conversion in argon, nitrogen, or vacuum. A quantitative model is derived from experiments for vacuum conversion to an uranium oxide-carbide kernel.« less

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

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

Magoulas, V; Charles Goergen, C; Ronald Oprea, R

2008-06-05

The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the productmore » throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.« less

Preliminary developments of MTR plates with uranium nitride

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

Durand, J.P.; Laudamy, P.; Richter, K.

1997-08-01

In the opinion of CERCA, the total weight of Uranium per MTR plate (without changing the external dimensions) cannot be further increased using U{sub 3}Si{sub 2}. Limits have been reached on plates with a thicker meat or loaded to 6g Ut/cm{sup 3}. The use of a denser fuel like Uranium mononitride could permit an increase in these limits. A collaboration between the Institute for Transuranium Elements (ITU), Joint Research Centre of the European Commission, and CERCA has been set ut. The preliminary studies at the ITU to check compatibility between aluminium and UN proved that there are no metallurgical interactionsmore » below 500{degrees}C. Feasibility of the manufacturing, on a laboratory scale at CERCA, of depleted Uranium mononitride plates loaded to 7 g Ut/cm{sup 3} has been demonstrated. The manufacturing process, however, is only one aspect of the development of a new fuel. The experience gained in the case of U{sub 3}Si{sub 2} has shown that the development of a new fuel requires considerable time and financial investment. Such a development certainly represents an effort of about 10 years.« less

Uranium mining in France

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

White, L.

Since the onset of the first ''oil shock'' in 1974, France has pursued a policy of steadily increasing energy independence based on nuclear power for generation of electricity. In 1973, nuclear reactors supplied only 8% of France's electrical power. A strong development effort lifted the nuclear share to 23% in 1980, to 66% in 1985, and the plan is to raise the total to 75% by 1990. In 1976, Cogema (Compagnie Generale des Matieres Nucleaires) was organized from the production division of France's Commissariat a l'Energie Atomique (CEA) to handle fuel supply and spent fuel reprocessing for the expanding industrymore » (see subsequent article on Cogema). In parallel with growth of the French nuclear power, Cogema has become a world leader in all aspects of the fuel cycle, providing services not only domestically but internationally as well. As a uranium mining company, Cogema has steadily developed domestic and foreign sources of supply, and over the years it has maintained the world's strongest uranium exploration effort throughout the ups and downs of the market. As a result, the company has become the world's leading uranium supplier, with about 20% of total production contributed either by its domestic mining divisions or overseas subsidiaries.« less

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

Boron coating on boron nitride coated nuclear fuels by chemical vapor deposition

NASA Astrophysics Data System (ADS)

Durmazuçar, Hasan H.; Gündüz, Güngör

2000-12-01

Uranium dioxide-only and uranium dioxide-gadolinium oxide (5% and 10%) ceramic nuclear fuel pellets which were already coated with boron nitride were coated with thin boron layer by chemical vapor deposition to increase the burn-up efficiency of the fuel during reactor operation. Coating was accomplished from the reaction of boron trichloride with hydrogen at 1250 K in a tube furnace, and then sintering at 1400 and 1525 K. The deposited boron was identified by infrared spectrum. The morphology of the coating was studied by using scanning electron microscope. The plate, grainy and string (fiber)-like boron structures were observed.

Peralkaline and peraluminous granites and related mineral deposits of the Arabian Shield, Kingdom of Saudi Arabia

USGS Publications Warehouse

Elliott, James E.

1983-01-01

Existing geochemical and geologic data for many parts of the Arabian Shield were compiled as a basis for evaluating the resource potential of the granites of the Shield. Commodities associated with granites that have potential for economic mineral deposits include tin, tungsten, molybdenum, beryllium, niobium, tantalum, zirconium, uranium, thorium, rare-earth elements, and fluorite. Prospecting methods useful in discriminating those granites having significant economic potential include reconnaissance geologic mapping, petrographic and mineralogic studies, geochemical sampling of rock and wadi sediment, and radiometric surveying.

Maps showing mineral resource assessment for copper and molybdenum in porphyry and stockwork deposits and for tungsten, iron, gold, copper, and silver in skarn deposits, Dillion 1 degree by 2 degrees Quadrangle, Idaho and Montana

USGS Publications Warehouse

Pearson, R.C.; Trautwein, C.M.; Moll, S.H.; Berger, B.R.; Hanna, W.F.; Loen, J.S.; Rowan, L.C.; Ruppel, E.T.; Segal, D.B.

1992-01-01

This report is one of several in the series that assess the mineral resources of the Dillon quadrangle. For the purpose of the assessment, mineral deposits· in the quadrangle that are either known or suspected from a knowledge of the geologic setting have been grou~d into 30 deposit types on the basis of mineralogy, commodity, or structural or depositional setting. The emphasis in these assessment reports is on metallic minerals, but some important nonmetallic minerals will also be assessed. Fossil fuels are beyond the scope of this investigation; phosphate and uranium have been investigated previously (Swanson, 1970; Wodzicki andKrason, 1981); and certain nonmetallic. minerals, including bulk commodities such as sand and gravel, are in large supply and thus not considered. The ·mineral resource assessment discussed in this report considers two deposit types: (1) porphyry or stockwork deposits of copper. and molybdenum (referred to generally in this report as porphyry deposits) and (2) skarn deposits of tungsten, iron, gold, copper, and silver. Combining copper and molybenum porphyry deposits into a single deposit type is believed necessary for this purpose· mainly because the two metals are found together in most .deposits in the quadrangle, a geochemical signature unique to each has not been determined, and the significant petrologic characteristics of many associated plutons are not well known, especially characterist~cs of subsurface plutons whose presence is .inferred from geophysical data. In assessing mineral resources,· we have adopted a general philosophy similar to that of Harrison and others (19S6). We attempt to identify those parts of the quadrangle that are favorable for the occurrence of mineral resources. We do not attempt to locate specific exploration targets nor to determine the quantity of reserves or resources present.

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)

SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

DOEpatents

Beaver, R.J.; Cherubini, J.H.

1963-05-14

A process is described for separating uranium from a nuclear fuel element comprising a uranium-containing core and a ferritic stainless steel clad by heating said element in a non-carburizing atmosphere at a temperature in the range 850-1050 un. Concent 85% C, rapidly cooling the heated element through the temperature range 815 un. Concent 85% to 650 EC to avoid annealing said steel, and then contacting the cooled element with an aqueous solution of nitric acid to selectively dissolve the uranium. (AEC)

Separation of uranium from technetium in recovery of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Friedman, H. A.

1984-06-01

A method for decontaminating uranium product from the Purex 5 process is described. Hydrazine is added to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO2(2+)) uranium and heptavalent technetius (TcO4-). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H2O2O4), and the Tc-oxalate complex is readily separated from the 10 uranium by solvent extraction with 30 vol % tributyl phosphate in n-dodecane.

Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium-Plutonium Carbide based fuel and 300 MWt Power

NASA Astrophysics Data System (ADS)

Clief Pattipawaej, Sandro; Su'ud, Zaki

2017-01-01

A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

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

Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber

The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactorsmore » with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).« less

Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

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

Lillo, Thomas; Rooyen, Isabella Van

2015-05-01

Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number ofmore » nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all grain boundaries and triple junctions contained precipitates with fission products and/or uranium, along with the differences in migration behavior between Pd, Ag and U, it was concluded that crystallographic grain boundary and triple junction parameters likely influence migration behavior.« less

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, II, William; Miller, Philip E.

1997-01-01

A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

DOEpatents

McLean, W. II; Miller, P.E.

1997-12-16

A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

Universal fuel basket for use with an improved oxide reduction vessel and electrorefiner vessel

DOEpatents

Herrmann, Steven D.; Mariani, Robert D.

2002-01-01

A basket, for use in the reduction of UO.sub.2 to uranium metal and in the electrorefining of uranium metal, having a continuous annulus between inner and outer perforated cylindrical walls, with a screen adjacent to each wall. A substantially solid bottom and top plate enclose the continuous annulus defining a fuel bed. A plurality of scrapers are mounted adjacent to the outer wall extending longitudinally thereof, and there is a mechanism enabling the basket to be transported remotely.

Fissioning uranium plasmas and nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Thom, K.

1975-01-01

Current research into uranium plasmas, gaseous-core (cavity) reactors, and nuclear-pumped lasers is discussed. Basic properties of fissioning uranium plasmas are summarized together with potential space and terrestrial applications of gaseous-core reactors and nuclear-pumped lasers. Conditions for criticality of a uranium plasma are outlined, and it is shown that the nonequilibrium state and the optical thinness of a fissioning plasma can be exploited for the direct conversion of fission fragment energy into coherent light (i.e., for nuclear-pumped lasers). Successful demonstrations of nuclear-pumped lasers are described together with gaseous-fuel reactor experiments using uranium hexafluoride.

In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

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

Haghighi, M. H.; Kring, C. T.; McGehee, J. T.

2002-02-26

The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 tomore » December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.« less

PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

DOEpatents

Moore, R.H.

1962-10-01

A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

A study on recovery of uranium in the anode basket residues delivered from the pyrochemical process of used nuclear fuel

NASA Astrophysics Data System (ADS)

Eun, H. C.; Kim, T. J.; Jang, J. H.; Kim, G. Y.; Park, S. B.; Yoon, D. S.; Kim, S. H.; Paek, S. W.; Lee, S. J.

2018-04-01

In this study, the chlorination of uranium oxide (UO2) using ammonium chloride and zirconium as chemical agents was conducted to recover the uranium in the anode basket residues from the pyrochemical process of used nuclear fuel. The chlorination of UO2 was predicted using thermodynamic equilibrium calculations. The experimental conditions for the chlorination were determined using a chlorination test with cerium oxide (CeO2). In the chlorination test, it was confirmed that UO2 was chlorinated into UCl3 at 320 °C, some UO2 remained without changes in the chemical form, and ZrO2, Zr2O, and ZrCl2 were generated as byproducts.

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-07-01

Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horiamore » Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.« less

United States-Gulf Cooperation Council Security Cooperation in a Multipolar World

DTIC Science & Technology

2014-10-01

including plu- tonium separation experiments, uranium enrichment and conversion experiments, and importing various uranium compounds.28 Subsequent...against political protest, a status shared with the two other remaining Arab monarchies, Morocco and Jordan . Geopolitically, the GCC as a region has...commitments, the UAE will not enrich uranium itself, relying instead on imported, enriched fuel. “Abu Dhabi Moves Ahead With Nuclear Program,” Middle

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

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

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required tomore » complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.« less

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

DOE PAGES

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; ...

2017-03-10

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Out-of-core Evaluations of Uranium Nitride-fueled Converters

NASA Technical Reports Server (NTRS)

Shimada, K.

1972-01-01

Two uranium nitride fueled converters were tested parametrically for their initial characterization and are currently being life-tested out of core. Test method being employed for the parametric and the diagnostic measurements during the life tests, and test results are presented. One converter with a rhenium emitter had an initial output power density of 6.9 W/ sq cm at the black body emitter temperature of 1900 K. The power density remained unchanged for the first 1000 hr of life test but degraded nearly 50% percent during the following 1000 hr. Electrode work function measurements indicated that the uranium fuel was diffusing out of the emitter clad of 0.635 mm. The other converter with a tungsten emitter had an initial output power density of 2.2 W/ sq cm at 1900 K with a power density of 3.9 W/sq cm at 4300 h. The power density suddenly degraded within 20 hr to practically zero output at 4735 hr.

75 FR 81675 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-28

... Fuel Cycle Facilities.'' FOR FURTHER INFORMATION CONTACT: Mekonen M. Bayssie, Regulatory Guide... Materials in Liquid and Gaseous Effluents from Nuclear Fuel Cycle Facilities,'' was published as Draft... guidance is applicable to nuclear fuel cycle facilities, with the exception of uranium milling facilities...

Impact of New Nuclear Data Libraries on Small Sized Long Life CANDLE HTGR Design Parameters

NASA Astrophysics Data System (ADS)

Liem, Peng Hong; Hartanto, Donny; Tran, Hoai Nam

2017-01-01

The impact of new evaluated nuclear data libraries (JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1) on the core characteristics of small-sized long-life CANDLE High Temperature Gas-Cooled Reactors (HTGRs) with uranium and thorium fuel cycles was investigated. The most important parameters of the CANDLE core characteristics investigated here covered (1) infinite multiplication factor of the fresh fuel containing burnable poison, (2) the effective multiplication factor of the equilibrium core, (3) the moving velocity of the burning region, (4) the attained discharge burnup, and (5) the maximum power density. The reference case was taken from the current JENDL-3.3 results. For the uranium fuel cycle, the impact of the new libraries was small, while significant impact was found for thorium fuel cycle. The findings indicated the needs of more accurate nuclear data libraries for nuclides involved in thorium fuel cycle in the future.

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Borowski, Stanley K.; Schnitzler, Bruce

2015-01-01

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

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

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

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

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Magnesium transport extraction of transuranium elements from LWR fuel

DOEpatents

Ackerman, John P.; Battles, James E.; Johnson, Terry R.; Miller, William E.; Pierce, R. Dean

1992-01-01

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid.

PubMed

Sung, Jinhyun; Kim, Jungsoo; Lee, Youngbae; Seol, Jeunggun; Ryu, Jaebong; Park, Kwangheon

2011-07-01

The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 °C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined.

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

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

BLENDING LOW ENRICHED URANIUM WITH DEPLETED URANIUM TO CREATE A SOURCE MATERIAL ORE THAT CAN BE PROCESSED FOR THE RECOVERY OF YELLOWCAKE AT A CONVENTIONAL URANIUM MILL

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

Schutt, Stephen M.; Hochstein, Ron F.; Frydenlund, David C.

2003-02-27

Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials atmore » a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.« less

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

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

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less