Novel fabrication of silicon carbide based ceramics for nuclear applications

NASA Astrophysics Data System (ADS)

Singh, Abhishek Kumar

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

Thermal barrier and support for nuclear reactor fuel core

DOEpatents

Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

1987-01-01

A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

Applications in the Nuclear Industry for Thermal Spray Amorphous Metal and Ceramic Coatings

NASA Astrophysics Data System (ADS)

Blink, J.; Farmer, J.; Choi, J.; Saw, C.

2009-06-01

Amorphous metal and ceramic thermal spray coatings have been developed with excellent corrosion resistance and neutron absorption. These coatings, with further development, could be cost-effective options to enhance the corrosion resistance of drip shields and waste packages, and limit nuclear criticality in canisters for the transportation, aging, and disposal of spent nuclear fuel. Iron-based amorphous metal formulations with chromium, molybdenum, and tungsten have shown the corrosion resistance believed to be necessary for such applications. Rare earth additions enable very low critical cooling rates to be achieved. The boron content of these materials and their stability at high neutron doses enable them to serve as high efficiency neutron absorbers for criticality control. Ceramic coatings may provide even greater corrosion resistance for waste package and drip shield applications, although the boron-containing amorphous metals are still favored for criticality control applications. These amorphous metal and ceramic materials have been produced as gas-atomized powders and applied as near full density, nonporous coatings with the high-velocity oxy-fuel process. This article summarizes the performance of these coatings as corrosion-resistant barriers and as neutron absorbers. This article also presents a simple cost model to quantify the economic benefits possible with these new materials.

Discrete Element Model for Simulations of Early-Life Thermal Fracturing Behaviors in Ceramic Nuclear Fuel Pellets

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

Hai Huang; Ben Spencer; Jason Hales

2014-10-01

A discrete element Model (DEM) representation of coupled solid mechanics/fracturing and heat conduction processes has been developed and applied to explicitly simulate the random initiations and subsequent propagations of interacting thermal cracks in a ceramic nuclear fuel pellet during initial rise to power and during power cycles. The DEM model clearly predicts realistic early-life crack patterns including both radial cracks and circumferential cracks. Simulation results clearly demonstrate the formation of radial cracks during the initial power rise, and formation of circumferential cracks as the power is ramped down. In these simulations, additional early-life power cycles do not lead to themore » formation of new thermal cracks. They do, however clearly indicate changes in the apertures of thermal cracks during later power cycles due to thermal expansion and shrinkage. The number of radial cracks increases with increasing power, which is consistent with the experimental observations.« less

Ceramics in nuclear waste management

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

Chikalla, T D; Mendel, J E

1979-05-01

Seventy-three papers are included, arranged under the following section headings: national programs for the disposal of radioactive wastes, waste from stability and characterization, glass processing, ceramic processing, ceramic and glass processing, leaching of waste materials, properties of nuclear waste forms, and immobilization of special radioactive wastes. Separate abstracts were prepared for all the papers. (DLC)

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, Xiangdong; Einziger, Robert E.

1997-01-01

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-08-12

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-01-28

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Glass/Ceramic Composites for Sealing Solid Oxide Fuel Cells

NASA Technical Reports Server (NTRS)

Bansal, Narottam P.; Choi, Sung R.

2007-01-01

A family of glass/ceramic composite materials has been investigated for use as sealants in planar solid oxide fuel cells. These materials are modified versions of a barium calcium aluminosilicate glass developed previously for the same purpose. The composition of the glass in mole percentages is 35BaO + 15CaO + 5Al2O3 + 10B2O3 + 35SiO2. The glass seal was found to be susceptible to cracking during thermal cycling of the fuel cells. The goal in formulating the glass/ ceramic composite materials was to (1) retain the physical and chemical advantages that led to the prior selection of the barium calcium aluminosilicate glass as the sealant while (2) increasing strength and fracture toughness so as to reduce the tendency toward cracking. Each of the composite formulations consists of the glass plus either of two ceramic reinforcements in a proportion between 0 and 30 mole percent. One of the ceramic reinforcements consists of alumina platelets; the other one consists of particles of yttria-stabilized zirconia wherein the yttria content is 3 mole percent (3YSZ). In preparation for experiments, panels of the glass/ceramic composites were hot-pressed and machined into test bars.

Spent Nuclear Fuel

EIA Publications

2015-01-01

Spent nuclear fuel data are collected by the U.S. Energy Information Administration (EIA) for the Department of Energy's Office of Standard Contract Management (Office of the General Counsel) on the Form GC-859, "Nuclear Fuel Data Survey." The data include detailed characteristics of spent nuclear fuel discharged from commercial U.S. nuclear power plants and currently stored at commercial sites in the United States. Utilities were not required to report spent nuclear fuel assemblies shipped to away-from-reactor, off-site facilities.

Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

NASA Astrophysics Data System (ADS)

Li, Bo-Shiuan

Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

DOEpatents

Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

2012-04-24

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

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

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

2011-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Characterisation of high temperature refractory ceramics for nuclear applications

NASA Astrophysics Data System (ADS)

Bottomley, P. D. W.; Wiss, Th; Janssen, A.; Cremer, B.; Thiele, H.; Manara, D.; Scheindlin, M.; Murray-Farthing, M.; Lajarge, P.; Menna, M.; Bouexière, D.; Rondinella, V. V.

2012-03-01

The ternary oxide ceramic system UO2-ZrO2-FeO is a refractory system that is of great relevance to the nuclear industry as it represents one of the main systems resulting from the interaction of the Zircaloy cladding, the UO2 fuel and the structural elements of a nuclear reactor. It is particularly the high temperature properties that require investigation; that is, when substantial overheating of the nuclear core occurs and interactions can lead to its degradation, melting and result in a severe nuclear accident. There has been much work on the UO2-ZrO2 system and also on the ternary system with FeO but there is still a need to examine 2 further aspects; firstly the effect of sub-oxidized systems, the UO2-Zr and FeO-Zr systems, and secondly the effect of Fe/Zr or Fe/U ratios on the melting point of the U-Zr-Fe oxide system. Samples of UO2-Zr and UO2-ZrO2-FeO were fabricated at ITU and then characterized by optical microscopy (OM) and X-ray diffraction to determine the ceramic's structure and verify the composition. Thereafter the samples are to be melted by laser flash heating and their liquidus and solidus temperatures determined by pyrometry. This programme is currently ongoing. The frozen samples, after testing, were then sectioned, polished and the molten zone micro-analytically examined by OM & SEM-EDS in order to determine its structure and composition and to compare with the existing phase diagrams. Examples of results from these systems will be given. Finally, a reacted Zr-FeO thermite mixture was examined, which had been used to generate high temperatures during tests of reactor melt-concrete interactions. The aim was to assess the reaction and estimate the heat generation from this novel technique. These results allow verification or improvement of the phase diagram and are of primary importance as input to models used to predict materials interactions in a severe nuclear accident.

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

NASA Technical Reports Server (NTRS)

Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

2011-01-01

Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

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.

Fully Ceramic Microencapsulated Fuel Development for LWR Applications

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

Snead, Lance Lewis; Besmann, Theodore M; Terrani, Kurt A

2012-01-01

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented includingmore » the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.« less

Nuclear fuel particles and method of making nuclear fuel compacts therefrom

DOEpatents

DeVelasco, Rubin I.; Adams, Charles C.

1991-01-01

Methods for making nuclear fuel compacts exhibiting low heavy metal contamination and fewer defective coatings following compact fabrication from a mixture of hardenable binder, such as petroleum pitch, and nuclear fuel particles having multiple layer fission-product-retentive coatings, with the dense outermost layer of the fission-product-retentive coating being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.3 g/cm.sup.3. Such particles can be pre-compacted in molds under relatively high pressures and then combined with a fluid binder which is ultimately carbonized to produce carbonaceous nuclear fuel compacts having relatively high fuel loadings.

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

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

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

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

Nuclear fuel cycle waste stream immobilization with cermets for improved thermal properties and waste consolidation

NASA Astrophysics Data System (ADS)

Ortega, Luis H.; Kaminski, Michael D.; Zeng, Zuotao; Cunnane, James

2013-07-01

In the pursuit of methods to improve nuclear waste form thermal properties and combine potential nuclear fuel cycle wastes, a bronze alloy was combined with an alkali, alkaline earth metal bearing ceramic to form a cermet. The alloy was prepared from copper and tin (10 mass%) powders. Pre-sintered ceramic consisting of cesium, strontium, barium and rubidium alumino-silicates was mixed with unalloyed bronze precursor powders and cold pressed to 300 × 103 kPa, then sintered at 600 °C and 800 °C under hydrogen. Cermets were also prepared that incorporated molybdenum, which has a limited solubility in glass, under similar conditions. The cermet thermal conductivities were seven times that of the ceramic alone. These improved thermal properties can reduce thermal gradients within the waste forms thus lowering internal temperature gradients and thermal stresses, allowing for larger waste forms and higher waste loadings. These benefits can reduce the total number of waste packages necessary to immobilize a given amount of high level waste and immobilize troublesome elements.

Anodes for protonic ceramic fuel cells (PCFCs) =

NASA Astrophysics Data System (ADS)

Nasani, Narendar

One of the more promising possibilities for future "green" electrical energy generation is the protonic ceramic fuel cell (PCFC). PCFCs offer a low-pollution technology to generate electricity electrochemically with high efficiency. Reducing the operating temperature of solid oxide fuel cells (SOFCs) to the 500-700°C range is desirable to reduce fabrication costs and improve overall longevity. This aim can be achieved by using protonic ceramic fuel cells (PCFCs) due to their higher electrolyte conductivity at these temperatures than traditional ceramic oxide-ion conducting membranes. This thesis deals with the state of the art Ni-BaZr0.85Y0.15O3-delta cermet anodes for PCFCs. The study of PCFCs is in its initial stage and currently only a few methods have been developed to prepare suitable anodes via solid state mechanical mixing of the relevant oxides or by combustion routes using nitrate precursors. This thesis aims to highlight the disadvantages of these traditional methods of anode preparation and to, instead, offer a novel, efficient and low cost nitrate free combustion route to prepare Ni-BaZr0.85Y0.15O3-delta cermet anodes for PCFCs. A wide range of techniques mainly X-ray diffraction (XRD), scanning electron microscopy (SEM), environmental scanning electron microscopy, (ESEM) and electrochemical impedance spectroscopy (EIS) were employed in the cermet anode study. The work also offers a fundamental examination of the effect of porosity, redox cycling behaviour, involvement of proton conducting oxide phase in PCFC cermet anodes and finally progresses to study the electrochemical performance of a state of the art anode supported PCFC. The polarisation behaviour of anodes has been assessed as a function of temperature (T), water vapour (pH2O), hydrogen partial pressures (pH2) and phase purity for electrodes of comparable microstructure. The impedance spectra generally show two arcs at high frequency R2 and low frequency R3 at 600 °C, which correspond to the

CERAMIC FUEL ELEMENT MATERIAL FOR A NEUTRONIC REACTOR AND METHOD OF FABRICATING SAME

DOEpatents

Duckworth, W.H.

1957-12-01

This patent relates to ceramic composition, and to neutronic reactor fuel elements formed therefrom. These ceramic elements have high density and excellent strength characteristics and are formed by conventional ceramic casting and sintering at a temperature of about 2700 deg F in a nitrogen atmosphere. The composition consists of silicon carbide, silicon, uranium oxide and a very small percentage of molybdenum. Compositions containing molybdenum are markedly stronger than those lacking this ingredient.

Swelling-resistant nuclear fuel

DOEpatents

Arsenlis, Athanasios [Hayward, CA; Satcher, Jr., Joe; Kucheyev, Sergei O [Oakland, CA

2011-12-27

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

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

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

1998-01-01

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Comprehensive Study on Ceramic Membranes for Low‐Cost Microbial Fuel Cells

PubMed Central

Pasternak, Grzegorz; Greenman, John

2016-01-01

Abstract Microbial fuel cells (MFCs) made with different types of ceramic membranes were investigated to find a low‐cost alternative to commercially available proton exchange membranes. The MFCs operated with fresh human urine as the fuel. Pyrophyllite and earthenware produced the best performance to reach power densities of 6.93 and 6.85 W m−3, respectively, whereas mullite and alumina achieved power densities of 4.98 and 2.60 W m−3, respectively. The results indicate the dependence of bio‐film growth and activity on the type of ceramic membrane applied. The most favourable conditions were created in earthenware MFCs. The performance of the ceramic membranes was related to their physical and chemical properties determined by environmental scanning electron microscopy and energy dispersive X‐ray spectroscopy. The cost of mullite, earthenware, pyrophyllite and alumina was estimated to be 13.61, 4.14, 387.96 and 177.03 GBP m−2, respectively. The results indicate that earthenware and mullite are good substitutes for commercially available proton exchange membranes, which makes the MFC technology accessible in developing countries. PMID:26692569

Implementation of focused ion beam (FIB) system in characterization of nuclear fuels and materials

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

A. Aitkaliyeva; J. W. Madden; B. D. Miller

2014-10-01

Beginning in 2007, a program was established at the Idaho National Laboratory to update key capabilities enabling microstructural and micro-chemical characterization of highly irradiated and/or radiologically contaminated nuclear fuels and materials at scales that previously had not been achieved for these types of materials. Such materials typically cannot be contact handled and pose unique hazards to instrument operators, facilities, and associated personnel. One of the first instruments to be acquired was a Dual Beam focused ion beam (FIB)-scanning electron microscope (SEM) to support preparation of transmission electron microscopy and atom probe tomography samples. Over the ensuing years, techniques have beenmore » developed and operational experience gained that has enabled significant advancement in the ability to characterize a variety of fuel types including metallic, ceramic, and coated particle fuels, obtaining insights into in-reactor degradation phenomena not obtainable by any other means. The following article describes insights gained, challenges encountered, and provides examples of unique results obtained in adapting Dual Beam FIB technology to nuclear fuels characterization.« less

Nuclear Fuels.

ERIC Educational Resources Information Center

Nash, J. Thomas

1983-01-01

Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

High temperature solid electrolyte fuel cell with ceramic electrodes

DOEpatents

Marchant, David D.; Bates, J. Lambert

1984-01-01

A solid oxide electrolyte fuel cell is described having a central electrolyte comprised of a HfO.sub.2 or ZrO.sub.2 ceramic stabilized and rendered ionically conductive by the addition of Ca, Mg, Y, La, Nd, Sm, Gd, Dy Er, or Yb. The electrolyte is sandwiched between porous electrodes of a HfO.sub.2 or ZrO.sub.2 ceramic stabilized by the addition of a rare earth and rendered electronically conductive by the addition of In.sub.2 O.sub.3. Alternatively, the anode electrode may be made of a metal such as Co, Ni, Ir Pt, or Pd.

High temperature solid electrolyte fuel cell with ceramic electrodes

DOEpatents

Bates, J.L.; Marchant, D.D.

A solid oxide electrolyte fuel cell is described having a central electrolyte comprised of a HfO/sub 2/ or ZrO/sub 2/ ceramic stabilized and rendered ionically conductive by the addition of Ca, Mg, Y, La, Nd, Sm, Gd, Dy Er, or Yb. The electrolyte is sandwiched between porous electrodes of a HfO/sub 2/ or ZrO/sub 2/ ceramic stabilized by the addition of a rare earth and rendered electronically conductive by the addition of In/sub 2/O/sub 3/. Alternatively, the anode electrode may be made of a metal such as Co, Ni, Ir Pt, or Pd.

Ceramic Integration Technologies for Advanced Energy Systems: Critical Needs, Technical Challenges, and Opportunities

NASA Technical Reports Server (NTRS)

Singh, Mrityunjay

2010-01-01

Advanced ceramic integration technologies dramatically impact the energy landscape due to wide scale application of ceramics in all aspects of alternative energy production, storage, distribution, conservation, and efficiency. Examples include fuel cells, thermoelectrics, photovoltaics, gas turbine propulsion systems, distribution and transmission systems based on superconductors, nuclear power generation and waste disposal. Ceramic integration technologies play a key role in fabrication and manufacturing of large and complex shaped parts with multifunctional properties. However, the development of robust and reliable integrated systems with optimum performance requires the understanding of many thermochemical and thermomechanical factors, particularly for high temperature applications. In this presentation, various needs, challenges, and opportunities in design, fabrication, and testing of integrated similar (ceramic ceramic) and dissimilar (ceramic metal) material www.nasa.gov 45 ceramic-ceramic-systems have been discussed. Experimental results for bonding and integration of SiC based Micro-Electro-Mechanical-Systems (MEMS) LDI fuel injector and advanced ceramics and composites for gas turbine applications are presented.

Nuclear fuels policy. Report of the Atlantic Council's Nuclear Fuels Policy Working Group

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

Not Available

1976-01-01

This Policy Paper recommends the actions deemed necessary to assure that future U.S. and non-Communist countries' nuclear fuels supply will be adequate, considering the following: estimates of modest growth in overall energy demand, electrical energy demand, and nuclear electrical energy demand in the U.S. and abroad, predicated upon the continuing trends involving conservation of energy, increased use of electricity, and moderate economic growth (Chap. I); possibilities for the development and use of all domestic resources providing energy alternatives to imported oil and gas, consonant with current environmental, health, and safety concerns (Chap. II); assessment of the traditional energy sources whichmore » provide current alternatives to nuclear energy (Chap. II); evaluation of realistic expectations for additional future energy supplies from prospective technologies: enhanced recovery from traditional sources and development and use of oil shales and synthetic fuels from coal, fusion and solar energy (Chap. II); an accounting of established nuclear technology in use today, in particular the light water reactor, used for generating electricity (Chap. III); an estimate of future nuclear technology, in particular the prospective fast breeder (Chap. IV); current and projected nuclear fuel demand and supply in the U.S. and abroad (Chaps. V and VI); the constraints encountered today in meeting nuclear fuels demand (Chap. VII); and the major unresolved issues and options in nuclear fuels supply and use (Chap. VIII). The principal conclusions and recommendations (Chap. IX) are that the U.S. and other industrialized countries should strive for increased flexibility of primary energy fuel sources, and that a balanced energy strategy therefore depends on the secure supply of energy resources and the ability to substitute one form of fuel for another.« less

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Method Of Making Closed End Ceramic Fuel Cell Tubes

DOEpatents

Borglum, Brian P.

2002-04-30

A method of manufacturing closed end ceramic fuel cell tubes with improved properties and higher manufacturing yield is disclosed. The method involves bonding an unfired cap to a hollow unfired tube to form a compound joint. The assembly is then fired to net shape without subsequent machining. The resultant closed end tube is superior in that it provides a leak-tight seal and its porosity is substantially identical to that of the tube wall. The higher manufacturing yield associated with the present method decreases overall fuel cell cost significantly.

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

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

Turner, John A; Clarno, Kevin T; Hansen, Glen A

2009-09-01

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied moremore » on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel

Ceramic Coatings for Corrosion Resistant Nuclear Waste Container Evaluated in Simulated Ground Water at 90?C

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

Haslam, J J; Farmer, J C

2004-03-31

Ceramic materials have been considered as corrosion resistant coatings for nuclear waste containers. Their suitability can be derived from the fully oxidized state for selected metal oxides. Several types of ceramic coatings applied to plain carbon steel substrates by thermal spray techniques have been exposed to 90 C simulated ground water for nearly 6 years. In some cases no apparent macroscopic damage such as coating spallation was observed in coatings. Thermal spray processes examined in this work included plasma spray, High Velocity Oxy Fuel (HVOF), and Detonation Gun. Some thermal spray coatings have demonstrated superior corrosion protection for the plainmore » carbon steel substrate. In particular the HVOF and Detonation Gun thermal spray processes produced coatings with low connected porosity, which limited the growth rate of corrosion products. It was also demonstrated that these coatings resisted spallation of the coating even when an intentional flaw (which allowed for corrosion of the carbon steel substrate underneath the ceramic coating) was placed in the coating. A model for prediction of the corrosion protection provided by ceramic coatings is presented. The model includes the effect of the morphology and amount of the porosity within the thermal spray coating and provides a prediction of the exposure time needed to produce a crack in the ceramic coating.« less

Comprehensive Study on Ceramic Membranes for Low-Cost Microbial Fuel Cells.

PubMed

Pasternak, Grzegorz; Greenman, John; Ieropoulos, Ioannis

2016-01-08

Microbial fuel cells (MFCs) made with different types of ceramic membranes were investigated to find a low-cost alternative to commercially available proton exchange membranes. The MFCs operated with fresh human urine as the fuel. Pyrophyllite and earthenware produced the best performance to reach power densities of 6.93 and 6.85 W m(-3), respectively, whereas mullite and alumina achieved power densities of 4.98 and 2.60 W m(-3), respectively. The results indicate the dependence of bio-film growth and activity on the type of ceramic membrane applied. The most favourable conditions were created in earthenware MFCs. The performance of the ceramic membranes was related to their physical and chemical properties determined by environmental scanning electron microscopy and energy dispersive X-ray spectroscopy. The cost of mullite, earthenware, pyrophyllite and alumina was estimated to be 13.61, 4.14, 387.96 and 177.03 GBP m(-2), respectively. The results indicate that earthenware and mullite are good substitutes for commercially available proton exchange membranes, which makes the MFC technology accessible in developing countries. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Vented nuclear fuel element

DOEpatents

Grossman, Leonard N.; Kaznoff, Alexis I.

1979-01-01

A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

A finite difference model used to predict the consolidation of a ceramic waste form produced from the electrometallurgical treatment of spent nuclear fuel.

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

Bateman, K. J.; Capson, D. D.

2004-03-29

Argonne National Laboratory (ANL) has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in the electrometallurgical treatment of spent Experimental Breeder Reactor-II (EBR-II) fuel. The ceramic waste process culminates with an elevated temperature operation. The processing conditions used by the furnace, for demonstration scale and production scale operations, are to be developed at Argonne National Laboratory-West (ANL-West). To assist in selecting the processing conditions of the furnace and to reduce the number of costly experiments, a finitemore » difference model was developed to predict the consolidation of the ceramic waste. The model accurately predicted the heating as well as the bulk density of the ceramic waste form. The methodology used to develop the computer model and a comparison of the analysis to experimental data is presented.« less

Ceramic Integration Technologies for Aerospace and Energy Systems: Technical Challenges and Opportunities

NASA Technical Reports Server (NTRS)

Singh, Mrityunjay

2007-01-01

Ceramic integration technology has been recognized as an enabling technology for the implementation of advanced ceramic systems in a number of high-temperature applications in aerospace, power generation, nuclear, chemical, and electronic industries. Various ceramic integration technologies (joining, brazing, attachments, repair, etc.) play a role in fabrication and manufacturing of large and complex shaped parts of various functionalities. However, the development of robust and reliable integrated systems with optimum performance requires the understanding of many thermochemical and thermomechanical factors, particularly for high temperature applications. In this presentation, various challenges and opportunities in design, fabrication, and testing of integrated similar (ceramic-ceramic) and dissimilar (ceramic-metal) material systems will be discussed. Experimental results for bonding and integration of SiC based LDI fuel injector, high conductivity C/C composite based heat rejection system, solid oxide fuel cells system, ultra high temperature ceramics for leading edges, and ceramic composites for thermostructural applications will be presented. Potential opportunities and need for the development of innovative design philosophies, approaches, and integrated system testing under simulated application conditions will also be discussed.

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco; ...

2017-07-19

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

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

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Method of making molten carbonate fuel cell ceramic matrix tape

DOEpatents

Maricle, Donald L.; Putnam, Gary C.; Stewart, Jr., Robert C.

1984-10-23

A method of making a thin, flexible, pliable matrix material for a molten carbonate fuel cell is described. The method comprises admixing particles inert in the molten carbonate environment with an organic polymer binder and ceramic particle. The composition is applied to a mold surface and dried, and the formed compliant matrix material removed.

A review into the use of ceramics in microbial fuel cells.

PubMed

Winfield, Jonathan; Gajda, Iwona; Greenman, John; Ieropoulos, Ioannis

2016-09-01

Microbial fuel cells (MFCs) offer great promise as a technology that can produce electricity whilst at the same time treat wastewater. Although significant progress has been made in recent years, the requirement for cheaper materials has prevented the technology from wider, out-of-the-lab, implementation. Recently, researchers have started using ceramics with encouraging results, suggesting that this inexpensive material might be the solution for propelling MFC technology towards real world applications. Studies have demonstrated that ceramics can provide stability, improve power and treatment efficiencies, create a better environment for the electro-active bacteria and contribute towards resource recovery. This review discusses progress to date using ceramics as (i) the structural material, (ii) the medium for ion exchange and (iii) the electrode for MFCs. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

Nuclear fuel pin scanner

DOEpatents

Bramblett, Richard L.; Preskitt, Charles A.

1987-03-03

Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

Spent nuclear fuel dry transfer system

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

Stewart, L.; Agace, S.

The U.S. Department of Energy is currently engaged in a cooperative program with the Electric Power Research Institute (EPRI) to design a spent nuclear fuel dry transfer system (DTS). The system will enable the transfer of individual spent nuclear fuel assemblies between a conventional top loading cask and multi-purpose canister in a shielded overpack, or accommodate spent nuclear fuel transfers between two conventional casks.

Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

NASA Astrophysics Data System (ADS)

Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.

2017-12-01

The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

Ion Selective Ceramics for Waste Separations. Input for Annual Accomplishments Report

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

Spoerke, Erik David

This report discusses“Ion-Selective Ceramics for Waste Separations” which aims to develop an electrochemical approach to remove fission product waste (e.g., Cs+ ) from the LiCl-KCl molten salts used in the pyroprocessing of spent nuclear fuel.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

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.

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Compositions and methods for treating nuclear fuel

DOEpatents

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2013-08-13

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Compositions and methods for treating nuclear fuel

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

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Microstructural changes in NiFe2O4 ceramics prepared with powders derived from different fuels in sol-gel auto-combustion technique

NASA Astrophysics Data System (ADS)

Chauhan, Lalita; Bokolia, Renuka; Sreenivas, K.

2016-05-01

Structural properties of Nickel ferrite (NiFe2O4) ceramics prepared from powders derived from sol gel auto-combustion method using different fuels (citric acid, glycine and Dl-alanine) are compared. Changes in the structural properties at different sintering temperatures are investigated. X-ray diffraction (XRD) confirms the formation of single phase material with cubic structure. Ceramics prepared using the different powders obtained from different fuels show that that there are no significant changes in lattice parameters. However increasing sintering temperatures show significant improvement in density and grain size. The DL-alanine fuel is found to be the most effective fuel for producing NIFe2O4 powders by the sol-gel auto combustion method and yields highly crystalline powders in the as-burnt stage itself at a low temperature (80 °C). Subsequent use of the powders in ceramic manufacturing produces dense NiFe2O4 ceramics with a uniform microstructure and a large grain size.

Used Nuclear Fuel: From Liability to Benefit

NASA Astrophysics Data System (ADS)

Orbach, Raymond L.

2011-03-01

Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

Environmental durability of ceramics and ceramic composites

NASA Technical Reports Server (NTRS)

Fox, Dennis S.

1992-01-01

An account is given of the current understanding of the environmental durability of both monolithic ceramics and ceramic-matrix composites, with a view to the prospective development of methods for the characterization, prediction, and improvement of ceramics' environmental durability. Attention is given to the environmental degradation behaviors of SiC, Si3N4, Al2O3, and glass-ceramic matrix compositions. The focus of corrosion prevention in Si-based ceramics such as SiC and Si3N4 is on the high and low sulfur fuel combustion-product effects encountered in heat engine applications of these ceramics; sintering additives and raw material impurities are noted to play a decisive role in ceramics' high temperature environmental response.

Zirconia-magnesia inert matrix fuel and waste form: Synthesis, characterization and chemical performance in an advanced fuel cycle

NASA Astrophysics Data System (ADS)

Holliday, Kiel Steven

There is a significant buildup in plutonium stockpiles throughout the world, because of spent nuclear fuel and the dismantling of weapons. The radiotoxicity of this material and proliferation risk has led to a desire for destroying excess plutonium. To do this effectively, it must be fissioned in a reactor as part of a uranium free fuel to eliminate the generation of more plutonium. This requires an inert matrix to volumetrically dilute the fissile plutonium. Zirconia-magnesia dual phase ceramic has been demonstrated to be a favorable material for this task. It is neutron transparent, zirconia is chemically robust, magnesia has good thermal conductivity and the ceramic has been calculated to conform to current economic and safety standards. This dissertation contributes to the knowledge of zirconia-magnesia as an inert matrix fuel to establish behavior of the material containing a fissile component. First, the zirconia-magnesia inert matrix is synthesized in a dual phase ceramic containing a fissile component and a burnable poison. The chemical constitution of the ceramic is then determined. Next, the material performance is assessed under conditions relevant to an advanced fuel cycle. Reactor conditions were assessed with high temperature, high pressure water. Various acid solutions were used in an effort to dissolve the material for reprocessing. The ceramic was also tested as a waste form under environmental conditions, should it go directly to a repository as a spent fuel. The applicability of zirconia-magnesia as an inert matrix fuel and waste form was tested and found to be a promising material for such applications.

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.

Carbide fuels for nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

Matthews, R. B.; Blair, H. T.; Chidester, K. M.; Davidson, K. V.; Stark, W. E.; Storms, E. K.

1991-09-01

A renewed interest in manned exploration of space has revitalized interest in the potential for advancing nuclear rocket technology developed during the 1960's. Carbide fuel performance, melting point, stability, fabricability and compatibility are key technology issues for advanced Nuclear Thermal Propulsion reactors. The Rover fuels development ended with proven carbide fuel forms with demonstrated operating temperatures up to 2700 K for over 100 minutes. The next generation of nuclear rockets will start where the Rover technology ended, but with a more rigorous set of operating requirements including operating lifetime to 10 hours, operating temperatures greater that 3000 K, low fission product release, and compatibility. A brief overview of Rover/NERVA carbide fuel development is presented. A new fuel form with the highest potential combination of operating temperature and lifetime is proposed that consists of a coated uranium carbide fuel sphere with built-in porosity to contain fission products. The particles are dispersed in a fiber reinforced ZrC matrix to increase thermal shock resistance.

NUCLEAR FUEL MATERIAL

DOEpatents

Goeddel, W.V.

1962-06-26

An improved method is given for making the carbides of nuclear fuel material. The metal of the fuel material, which may be a fissile and/or fertile material, is transformed into a silicide, after which the silicide is comminuted to the desired particle size. This silicide is then carburized at an elevated temperature, either above or below the melting point of the silicide, to produce an intimate mixture of the carbide of the fuel material and the carbide of silicon. This mixture of the fuel material carbide and the silicon carbide is relatively stable in the presence of moisture and does not exhibit the highly reactive surface condition which is observed with fuel material carbides made by most other known methods. (AEC)

Nuclear power generation and fuel cycle report 1996

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

NONE

1996-10-01

This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

Gaseous fuel nuclear reactor research

NASA Technical Reports Server (NTRS)

Schwenk, F. C.; Thom, K.

1975-01-01

Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

Nuclear fuel element

DOEpatents

Zocher, Roy W.

1991-01-01

A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

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

Marra, J.

2010-09-29

Rising global energy demands coupled with increased environmental concerns point to one solution; they must reduce their dependence on fossil fuels that emit greenhouse gases. As the global community faces the challenge of maintaining sovereign nation security, reducing greenhouse gases, and addressing climate change nuclear power will play a significant and likely growing role. In the US, nuclear energy already provides approximately one-fifth of the electricity used to power factories, offices, homes, and schools with 104 operating nuclear power plants, located at 65 sites in 31 states. Additionally, 19 utilities have applied to the US Nuclear Regulatory Commission (NRC) formore » construction and operating licenses for 26 new reactors at 17 sites. This planned growth of nuclear power is occurring worldwide and has been termed the 'nuclear renaissance.' As major industrial nations craft their energy future, there are several important factors that must be considered about nuclear energy: (1) it has been proven over the last 40 years to be safe, reliable and affordable (good for Economic Security); (2) its technology and fuel can be domestically produced or obtained from allied nations (good for Energy Security); and (3) it is nearly free of greenhouse gas emissions (good for Environmental Security). Already an important part of worldwide energy security via electricity generation, nuclear energy can also potentially play an important role in industrial processes and supporting the nation's transportation sector. Coal-to-liquid processes, the generation of hydrogen and supporting the growing potential for a greatly increased electric transportation system (i.e. cars and trains) mean that nuclear energy could see dramatic growth in the near future as we seek to meet our growing demand for energy in cleaner, more secure ways. In order to address some of the prominent issues associated with nuclear power generation (i.e., high capital costs, waste management, and

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

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

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input

Nuclear fuel in a reactor accident.

PubMed

Burns, Peter C; Ewing, Rodney C; Navrotsky, Alexandra

2012-03-09

Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

Corrosion Behavior and Microstructure Influence of Glass-Ceramic Nuclear Waste Forms

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

Matthew Asmussen, R.; Neeway, James J.; Kaspar, Tiffany C.

Glass ceramic waste forms present a potentially viable technology for the long term immobilization and disposal of liquid nuclear wastes. Through control of chemistry during fabrication, such waste forms can have designed secondary crystalline phases within a borosilicate glass matrix. In this work, a glass ceramic containing powellite and oxyapatite secondary phases was tested for its corrosion properties in dilute conditions using single pass flow through testing (SPFT). Three glass ceramic samples were prepared using different cooling rates to produce samples with varying microstructure sizes. In testing at 90 °C in buffered pH 7 and pH 9 solutions, it wasmore » found that increasing pH and decreasing microstructure size (resulting from rapid cooling during fabrication) both led to a reduction in overall corrosion rate. The phases of the glass ceramic were found, using a combination of solutions analysis, SEM and AFM, to corrode preferably in the order of powellite > bulk glass matrix > oxyapatite.« less

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.

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

UN TRISO Compaction in SiC for FCM Fuel Irradiations

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

Terrani, Kurt A.; Trammell, Michael P.; Kiggans, James O.

2016-11-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is conducting research and development to elevate the technology readiness level of Fully Ceramic Microencapsulated (FCM) fuels, a candidate nuclear fuel with potentially enhanced accident tolerance due to very high fission product retention. One of the early activities in FY17 was to demonstrate production of FCM pellets with uranium nitride TRISO particles. This was carried out in preparation of the larger pellet production campaign in support of the upcoming irradiation testing of this fuel form at INL’s Advanced Test Reactor.

Nuclear Fuel Cycle Options Catalog: FY16 Improvements and Additions

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

Price, Laura L.; Barela, Amanda Crystal; Schetnan, Richard Reed

2016-08-31

The United States Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Program sponsors nuclear fuel cycle research and development. As part of its Fuel Cycle Options campaign, the DOE has established the Nuclear Fuel Cycle Options Catalog. The catalog is intended for use by the Fuel Cycle Technologies Program in planning its research and development activities and disseminating information regarding nuclear energy to interested parties. The purpose of this report is to document the improvements and additions that have been made to the Nuclear Fuel Cycle Options Catalog in the 2016 fiscal year.

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

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

NONE

2013-07-01

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

Formation of thin walled ceramic solid oxide fuel cells

DOEpatents

Claar, Terry D.; Busch, Donald E.; Picciolo, John J.

1989-01-01

To reduce thermal stress and improve bonding in a high temperature monolithic solid oxide fuel cell (SOFC), intermediate layers are provided between the SOFC's electrodes and electrolyte which are of different compositions. The intermediate layers are comprised of a blend of some of the materials used in the electrode and electrolyte compositions. Particle size is controlled to reduce problems involving differential shrinkage rates of the various layers when the entire structure is fired at a single temperature, while pore formers are provided in the electrolyte layers to be removed during firing for the formation of desired pores in the electrode layers. Each layer includes a binder in the form of a thermosetting acrylic which during initial processing is cured to provide a self-supporting structure with the ceramic components in the green state. A self-supporting corrugated structure is thus formed prior to firing, which the organic components of the binder and plasticizer removed during firing to provide a high strength, high temperature resistant ceramic structure of low weight and density.

SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS

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

Matthew C. Morrison; Kenneth J. Bateman; Michael F. Simpson

2010-11-01

ABSTRACT SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS Matthew C. Morrison, Kenneth J. Bateman, Michael F. Simpson Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 The ceramic waste process is the intended method for disposing of waste salt electrolyte, which contains fission products from the fuel-processing electrorefiners (ER) at the INL. When mixed and processed with other materials, the waste salt can be stored in a durable ceramic waste form (CWF). The development of the CWF has recently progressed from small-scale testing and characterization to full-scale implementation and experimentation using surrogate materialsmore » in lieu of the ER electrolyte. Two full-scale (378 kg and 383 kg) CWF test runs have been successfully completed with final densities of 2.2 g/cm3 and 2.1 g/cm3, respectively. The purpose of the first CWF was to establish material preparation parameters. The emphasis of the second pre-qualification test run was to evaluate a preliminary multi-section CWF container design. Other considerations were to finalize material preparation parameters, measure the material height as it consolidates in the furnace, and identify when cracking occurs during the CWF cooldown process.« less

International nuclear fuel cycle fact book. Revision 4

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

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries -more » a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.« less

International Nuclear Fuel Cycle Fact Book. Revision 5

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

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries -more » a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.« less

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

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

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

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.

Natural Silica Sand/Alumina Ceramic Composites: Promising Candidates for Fuel-Cell Sealants

NASA Astrophysics Data System (ADS)

Hidayat, N.; Istiqomah; Widianto, M. Y. H.; Taufiq, A.; Sunaryono; Triwikantoro; Zainuri, M.; Baqiya, M. A.; Aristia, G.; Pratapa, S.

2017-05-01

An attempt has been developed to establish the prospect of the useful application of Indonesian natural silica sand, instead of commercially expensive materials, as a future fuel-cell sealant. The sand was initially washed and ball-milled at 150 rpm for 60 minutes and then heated at 1000 °C for the same duration. The resulting powder was then mixed with alumina powder at various amounts and shaped into discs before sintering at 1150 °C and 1250 °C to produce compact ceramics. The diameter shrinkage, porosity, and density of the ceramics were evaluated by Archimedes method. Their crystalline phase composition was quantified by Rietveld refinement analysis on the X-ray diffraction (XRD) data and the phase weight fraction was then used for coefficient of thermal expansion (CTE) evaluation. It was observed that the bulk density increased while the porosity decreased with alumina addition. The XRD data analysis revealed that the prepared silica sand contains a very high purity of quartz-SiO2, i.e. 97.8(18)%. The sintering temperatures of 1150 °C and 1250 °C transformed some quartz-SiO2 to crystobalite-SiO2. All the calcite-CaCO3 exhibited reaction sintering with SiO2 forming wollastonite-CaSiO3. Therefore, the ceramic composites contained SiO2/Al2O3/CaSiO3. Regarding CTE, all of the composites meet the criteria for fuel-cell sealants, in the range of 9-12 ppm/°C.

Cosmic ray muons for spent nuclear fuel monitoring

NASA Astrophysics Data System (ADS)

Chatzidakis, Stylianos

There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks

Microstructural changes in NiFe{sub 2}O{sub 4} ceramics prepared with powders derived from different fuels in sol-gel auto-combustion technique

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

Chauhan, Lalita, E-mail: chauhan.lalita5@gmail.com; Sreenivas, K.; Bokolia, Renuka

2016-05-23

Structural properties of Nickel ferrite (NiFe{sub 2}O{sub 4}) ceramics prepared from powders derived from sol gel auto-combustion method using different fuels (citric acid, glycine and Dl-alanine) are compared. Changes in the structural properties at different sintering temperatures are investigated. X-ray diffraction (XRD) confirms the formation of single phase material with cubic structure. Ceramics prepared using the different powders obtained from different fuels show that that there are no significant changes in lattice parameters. However increasing sintering temperatures show significant improvement in density and grain size. The DL-alanine fuel is found to be the most effective fuel for producing NIFe{sub 2}O{submore » 4} powders by the sol-gel auto combustion method and yields highly crystalline powders in the as-burnt stage itself at a low temperature (80 °C). Subsequent use of the powders in ceramic manufacturing produces dense NiFe{sub 2}O{sub 4} ceramics with a uniform microstructure and a large grain size.« less

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, Darrell F.

1993-01-01

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Method and apparatus for close packing of nuclear fuel assemblies

DOEpatents

Newman, D.F.

1993-03-30

The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

Review of palm oil fuel ash and ceramic waste in the production of concrete

NASA Astrophysics Data System (ADS)

Natasya Mazenan, Puteri; Sheikh Khalid, Faisal; Shahidan, Shahiron; Shamsuddin, Shamrul-mar

2017-11-01

High demand for cement in the concrete production has been increased which become the problems in the industry. Thus, this problem will increase the production cost of construction material and the demand for affordable houses. Moreover, the production of Portland cement leads to the release of a significant amount of CO2 and other gases leading to the effect on global warming. The need for a sustainable and green construction building material is required in the construction industry. Hence, this paper presents utilization of palm oil fuel ash and ceramic waste as partial cement replacement in the production of concrete. Using both of this waste in the concrete production would benefit in many ways. It is able to save cost and energy other than protecting the environment. In short, 20% usage of palm oil fuel ash and 30% replacement of ceramic waste as cement replacement show the acceptable and satisfactory strength of concrete.

OECD/NEA Ongoing activities related to the nuclear fuel cycle

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

Cornet, S.M.; McCarthy, K.; Chauvin, N.

2013-07-01

As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclearmore » systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)« less

Spent Nuclear Fuel Transport Reliability Study

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

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

This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

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

DOEpatents

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

2010-02-23

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

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

DOEpatents

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

2013-09-03

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

Development of Nano-crystalline Doped-Ceramic Enabled Fiber Sensors for High Temperature In-Situ Monitoring of Fossil Fuel Gases

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

Xiao, Hai; Dong, Junhang; Lin, Jerry

2012-03-01

This is a final technical report for the first project year from July 1, 2005 to Jan 31, 2012 for DoE/NETL funded project DE-FC26-05NT42439: Development of Nanocrystalline Doped-Ceramic Enabled Fiber Sensors for High Temperature In-Situ Monitoring of Fossil Fuel Gases. This report summarizes the technical progresses and achievements towards the development of novel nanocrystalline doped ceramic material-enabled optical fiber sensors for in situ and real time monitoring the gas composition of flue or hot gas streams involved in fossil-fuel based power generation and hydrogen production.

Tubular solid oxide fuel cells with porous metal supports and ceramic interconnections

DOEpatents

Huang, Kevin [Export, PA; Ruka, Roswell J [Pittsburgh, PA

2012-05-08

An intermediate temperature solid oxide fuel cell structure capable of operating at from 600.degree. C. to 800.degree. C. having a very thin porous hollow elongated metallic support tube having a thickness from 0.10 mm to 1.0 mm, preferably 0.10 mm to 0.35 mm, a porosity of from 25 vol. % to 50 vol. % and a tensile strength from 700 GPa to 900 GPa, which metallic tube supports a reduced thickness air electrode having a thickness from 0.010 mm to 0.2 mm, a solid oxide electrolyte, a cermet fuel electrode, a ceramic interconnection and an electrically conductive cell to cell contact layer.

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

International nuclear fuel cycle fact book. Revision 6

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

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

1986-01-01

The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

Nuclear Fuels & Materials Spotlight Volume 5

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

Petti, David Andrew

2016-10-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system.more » • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.« less

Spent Nuclear Fuel (SNF) Project Execution Plan

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

LEROY, P.G.

2000-11-03

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

Nuclear Fuel Cycle Introductory Concepts

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

Karpius, Peter Joseph

2017-02-02

The nuclear fuel cycle is a complex entity, with many stages and possibilities, encompassing natural resources, energy, science, commerce, and security, involving a host of nations around the world. This overview describes the process for generating nuclear power using fissionable nuclei.

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…

Composite construction for nuclear fuel containers

DOEpatents

Cheng, Bo-Ching [Fremont, CA; Rosenbaum, Herman S [Fremont, CA; Armijo, Joseph S [Saratoga, CA

1987-01-01

An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

NASA Astrophysics Data System (ADS)

Ortega, Luis Humberto

The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE) program, that has been investigating technologies to improve fuel cycle sustainability and proliferation resistance. One of the program's goals is to reduce the amount of radioactive waste requiring repository disposal. Cesium and strontium are two primary heat sources during the first 300 years of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of these isotopes from spent nuclear fuel will reduce the activity of the bulk spent fuel, reducing the heat given off by the waste. Once the cesium and strontium are separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized. This study is focused on a method to immobilize a cesium- and strontium-bearing radioactive liquid waste stream. While there are various schemes to remove these isotopes from spent fuel, this study has focused on a nitric acid based liquid waste. The waste liquid was mixed with the bentonite, dried then sintered. To be effective sintering temperatures from 1100 to 1200°C were required, and waste concentrations must be at least 25 wt%. The product is a leach resistant ceramic solid with the waste elements embedded within alumino-silicates and a silicon rich phase. The cesium is primarily incorporated into pollucite and the strontium into a monoclinic feldspar. The simulated waste was prepared from nitrate salts of stable ions. These ions were limited to cesium, strontium, barium and rubidium. Barium and rubidium will be co-extracted during separation due to similar chemical properties to cesium and strontium. The waste liquid was added to the bentonite clay incrementally with drying steps between each addition. The dry powder was pressed and then sintered at various temperatures. The maximum loading tested is 32 wt. percent waste, which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent strontium, and 2.0 wt. percent rubidium. Lower loadings of waste

An analysis of international nuclear fuel supply options

NASA Astrophysics Data System (ADS)

Taylor, J'tia Patrice

As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

DTIC Science & Technology

2007-11-01

critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation policy discouraged commercial nuclear fuel...perhaps the most critical question in this decade for strengthening the nuclear nonproliferation regime: how can access to sensitive fuel cycle...process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of uranium

Composite construction for nuclear fuel containers

DOEpatents

Cheng, B. C.; Rosenbaum, H. S.; Armijo, J. S.

1987-04-21

Disclosed is an improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof. 1 fig.

Grooved Fuel Rings for Nuclear Thermal Rocket Engines

NASA Technical Reports Server (NTRS)

Emrich, William

2009-01-01

An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

Separation of the rare-earth fission product poisons from spent nuclear fuel

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

Christian, Jerry D.; Sterbentz, James W.

A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2more » in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.« less

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%.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-16

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

NASA Astrophysics Data System (ADS)

Vetter, M.; Wang, J.

2017-12-01

The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

Ceramic plasma-sprayed coating of melting crucibles for casting metal fuel slugs

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

K.H. Kim; C.T. Lee; C.B. Lee

2013-10-01

Thermal cycling and melt reaction studies of ceramic coatings plasma-sprayed on Nb substrates were carried out to evaluate the performance of barrier coatings for metallic fuel casting applications. Thermal cycling tests of the ceramic plasma-sprayed coatings to 1450 degrees C showed that HfN, TiC, ZrC, and Y2O3 coating had good cycling characteristics with few interconnected cracks even after 20 cycles. Interaction studies by 1550 degrees C melt dipping tests of the plasma-sprayed coatings also indicated that HfN and Y2O3 do not form significant reaction layer between U–20 wt.% Zr melt and the coating layer. Plasma-sprayed Y2O3 coating exhibited the mostmore » promising characteristics among HfN, TiC, ZrC, and Y2O3 coating.« less

Process for making ultra-fine ceramic particles

DOEpatents

Stangle, Gregory C.; Venkatachari, Koththavasal R.; Ostrander, Steven P.; Schulze, Walter A.

1995-01-01

A process for producing ultra-fine ceramic particles in which droplets are formed from a ceramic precursor mixture containing a metal cation, a nitrogen-containing fuel, a solvent, and an anion capable of participating in an anionic oxidation-reduction reaction with the nitrogen containing fuel. The nitrogen-containing fuel contains at least three nitrogen atoms, at least one oxygen atom, and at least one carbon atom. The ceramic precursor mixture is dried to remove at least 85 weight percent of the solvent, and the dried mixture is then ignited to form a combusted powder.

Pyrolytic carbon-coated nuclear fuel

DOEpatents

Lindemer, Terrence B.; Long, Jr., Ernest L.; Beatty, Ronald L.

1978-01-01

An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.

Nuclear fuel electrorefiner

DOEpatents

Ahluwalia, Rajesh K.; Hua, Thanh Q.

2004-02-10

The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

Alkaline fuel cell: carbon nanobeads coated with metal catalyst over porous ceramic for hydrogen electrode

NASA Astrophysics Data System (ADS)

Chatterjee, A. K.; Sharon, Maheshwar; Banerjee, Rangan

The development of a hydrogen electrode using a porous ceramic coated with carbon nanobeads for an alkaline fuel cell (AFC) is reported. This electrode can provide necessary strength and porosity to enable hydrogen to diffuse without allowing electrolyte to percolate inside the electrode. Various catalysts (Pt, Ni, Co and Fe) are electrochemically dispersed over the carbon nanobeads to examine their performance in the alkaline fuel cell. Turpentine oil has been used as a precursor for preparing the carbon nanobeads by a chemical vapour deposition technique. Scanning electron microscopic and transmission electron microscopic images show that the carbon nanobeads have sizes between 500 and 650 nm and are spread uniformly over the entire ceramic substrate. X-ray diffraction (XRD) patterns indicate that the nanobeads are graphitic in nature. Thus, the electrode is highly conductive. The current-voltage characteristics and chronopotentiometry of a half cell (i.e. hydrogen electrode coated with different electrocatalysts) and a full cell (using both hydrogen and oxygen electrodes) with 30% KOH solution are measured. About 93% of the theoretical hydrogen dissociation voltage is obtained with Ni and Pt catalyst. All other metals (Co and Fe) give a lower voltage. Ni-coated carbon nanobeads deposited over a ceramic oxide can be used in place of Raney nickel electrode as their characteristics are similar to those of a platinum electrode.

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Letter report on PCT/Monolith glass ceramic corrosion tests

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

Crawford, Charles L.

2015-09-24

The Savannah River National Laboratory (SRNL) is collaborating with personnel from Pacific Northwest National Laboratory (PNNL) to study advanced waste form glass ceramics for immobilization of waste from Used Nuclear Fuel (UNF) separations processes. The glass ceramic waste forms take advantage of both crystalline and glassy phases where ‘troublesome’ elements (e.g., low solubility in glass or very long-lived) partition to highly durable ceramic phases with the remainder of elements residing in the glassy phase. The ceramic phases are tailored to create certain minerals or unique crystalline structures that can host the radionuclides by binding them in their specific crystalline networkmore » while not adversely impacting the residual glass network (Crum et al., 2011). Glass ceramics have been demonstrated using a scaled melter test performed in a pilot scale (1/4 scale) cold crucible induction melter (CCIM) (Crum et al., 2014; Maio et al., 2015). This report summarizes recent results from both Phase I and Phase II bench scale tests involving crucible fabrication and corrosion testing of glass ceramics using the Product Consistency Test (PCT). Preliminary results from both Phase I and Phase II bench scale tests involving statistically designed matrices have previously been reported (Crawford, 2013; Crawford, 2014).« less

SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

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

Marshall, Douglas W.

High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must havemore » very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

Antineutrino Monitoring of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Brdar, Vedran; Huber, Patrick; Kopp, Joachim

2017-11-01

Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries worldwide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this paper, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear-waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to reverify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in contaminated areas such as the Hanford site in Washington state.

Nuclear Energy and Synthetic Liquid Transportation Fuels

NASA Astrophysics Data System (ADS)

McDonald, Richard

2012-10-01

This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

FUEL ELEMENTS FOR NUCLEAR REACTORS

DOEpatents

Blainey, A.; Lloyd, H.

1961-07-11

A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

Composite nuclear fuel fabrication methodology for gas fast reactors

NASA Astrophysics Data System (ADS)

Vasudevamurthy, Gokul

An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

78 FR 56775 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

... radiological impacts of spent nuclear fuel and high-level waste disposal. DATES: Submit comments on the... determination. The ``Offsite radiological impacts of spent nuclear fuel and high-level waste disposal'' issue.... Geologic Repository--Technical Feasibility and Availability C3. Storage of Spent Nuclear Fuel C3.a...

Thermoacoustic enhancements for nuclear fuel rods and other high temperature applications

DOEpatents

Garrett, Steven L.; Smith, James A.; Kotter, Dale K.

2017-05-09

A nuclear thermoacoustic device includes a housing defining an interior chamber and a portion of nuclear fuel disposed in the interior chamber. A stack is disposed in the interior chamber and has a hot end and a cold end. The stack is spaced from the portion of nuclear fuel with the hot end directed toward the portion of nuclear fuel. The stack and portion of nuclear fuel are positioned such that an acoustic standing wave is produced in the interior chamber. A frequency of the acoustic standing wave depends on a temperature in the interior chamber.

Nuclear reactor composite fuel assembly

DOEpatents

Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

1980-01-01

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Nuclear reactor fuel element

DOEpatents

Johnson, Carl E.; Crouthamel, Carl E.

1980-01-01

A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

Monitoring arrangement for vented nuclear fuel elements

DOEpatents

Campana, Robert J.

1981-01-01

In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

Fuel Fabrication and Nuclear Reactors

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

Karpius, Peter Joseph

2017-02-02

The uranium from the enrichment plant is still in the form of UF 6. UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

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.

77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-30

... DEPARTMENT OF ENERGY Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel Cycle Technologies, Office of Nuclear Energy, Department of Energy. ACTION: Notice of meeting. SUMMARY: The Office of Fuel Cycle Technologies will be hosting a one- day informational meeting at the Argonne...

Variants of closing the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

2015-12-01

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

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

Weber, William J.; Zhang, Yanwen

This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effectsmore » of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.« less

The myth of the ``proliferation-resistant'' closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2000-07-01

National nuclear energy programs that engage in reprocessing of spent nuclear fuel (SNF) and the development of "closed" nuclear fuel cycles based on the utilization of plutonium process and store large quantities of weapons-usable nuclear materials in forms vulnerable to diversion or theft by national or subnational groups. Proliferation resistance, an idea dating back at least as far as the International Fuel Cycle Evaluation (INFCE) of the late 1970s, is a loosely defined term referring to processes for chemical separation of SNF that do not extract weapons-usable materials in a purified form.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

Recent Progress in Using Advanced Characterization and Modeling Approaches to Study Radiation Effects in Oxide Ceramics

DOE PAGES

Bai, Xian-Ming

2014-10-23

I serve as a Guest Editor for the Nuclear Materials Committee of the TMS Structural Materials Division, and coordinated the topic ‘‘Radiation Effects in Oxide Ceramics and Novel LWR Fuels" for JOM in the December 2014 issue. I selected five articles related this topic. These articles talk about some recent progress of using advanced experimental and modeling tools to study radiation effects in oxide ceramics at atomistic scale and mesoscale. In this guest editor commentary article, I summarize the novel aspects of these papers and also provide some suggestions for future research directions.

Calcium titanium silicate based glass-ceramic for nuclear waste immobilisation

NASA Astrophysics Data System (ADS)

Sharma, K.; Srivastav, A. P.; Goswami, M.; Krishnan, Madangopal

2018-04-01

Titanate based ceramics (synroc) have been studied for immobilisation of nuclear wastes due to their high radiation and thermal stability. The aim of this study is to synthesis glass-ceramic with stable phases from alumino silicate glass composition and study the loading behavior of actinides in glass-ceramics. The effects of CaO and TiO2 addition on phase evolution and structural properties of alumino silicate based glasses with nominal composition x(10CaO-9TiO2)-y(10Na2O-5 Al2O3-56SiO2-10B2O3); where z = x/y = 1.4-1.8 are reported. The glasses are prepared by melt-quench technique and characterized for thermal and structural properties using DTA and Raman Spectroscopy. Glass transition and peak crystallization temperatures decrease with increase of CaO and TiO2 content, which implies the weakening of glass network and increased tendency of glasses towards crystallization. Sphene (CaTiSiO5) and perovskite (CaTiO3) crystalline phases are confirmed from XRD which are well known stable phase for conditioning of actinides. The microsturcture and elemental analysis indicate the presence of actinide in stable crystalline phases.

Laser surface treatment of porous ceramic substrate for application in solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Mahmod, D. S. A.; Khan, A. A.; Munot, M. A.; Glandut, N.; Labbe, J. C.

2016-08-01

Laser has offered a large number of benefits for surface treatment of ceramics due to possibility of localized heating, very high heating/cooling rates and possibility of growth of structural configurations only produced under non-equilibrium high temperature conditions. The present work investigates oxidation of porous ZrB2-SiC sintered ceramic substrates through treatment by a 1072 ± 10 nm ytterbium fiber laser. A multi-layer structure is hence produced showing successively oxygen rich distinct layers. The porous bulk beneath these layers remained unaffected as this laser-formed oxide scale and protected the substrate from oxidation. A glassy SiO2 structure thus obtained on the surface of the substrate becomes subject of interest for further research, specifically for its utilization as solid protonic conductor in Solid Oxide Fuel Cells (SOFCs).

Compliant fuel cell system

DOEpatents

Bourgeois, Richard Scott [Albany, NY; Gudlavalleti, Sauri [Albany, NY

2009-12-15

A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

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.

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

Effect of binder burnout on the sealing performance of glass ceramics for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Ertugrul, Tugrul Y.; Celik, Selahattin; Mat, Mahmut D.

2013-11-01

The glass ceramics composite sealants are among few materials suitable for the solid oxide fuel cells (SOFC) due to their high operating temperatures (600 °C-850 °C). The glass ceramics chemically bond to both the metallic interconnector and the ceramic electrolyte and provide a gas tight connection. A careful and several stages manufacturing procedure is required to obtain a gas tight sealing. In this study, effects of binder burnout process on the sealing performance are investigated employing commercially available glass ceramic powders. The glass ceramic laminates are produced by mixing glass ceramic powders with the organic binders and employing a tape casting method. The laminates are sandwiched between the metallic interconnectors of an SOFC cell. The burnout and subsequent sealing quality are analyzed by measuring leakage rate and final macrostructure of sealing region. The effects of heating rate, dead weight load, solid loading, carrier gas and their flow rates are investigated. It is found that sealing quality is affected from all investigated parameters. While a slower heating rate is required for a better burnout, the mass flow rate of sweep gas must be adequate for removal of the burned gas. The leakage rate is reduced to 0.1 ml min-1 with 2 °C min-1 + 1 °C min-1 heating rate, 86.25% solid loading, 200 N dead weight load and 500 ml min-1 sweep gas flow rate.

Hanford Spent Nuclear Fuel Project recommended path forward

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

Fulton, J.C.

The Spent Nuclear Fuel Project (the Project), in conjunction with the U.S. Department of Energy-commissioned Independent Technical Assessment (ITA) team, has developed engineered alternatives for expedited removal of spent nuclear fuel, including sludge, from the K Basins at Hanford. These alternatives, along with a foreign processing alternative offered by British Nuclear Fuels Limited (BNFL), were extensively reviewed and evaluated. Based on these evaluations, a Westinghouse Hanford Company (WHC) Recommended Path Forward for K Basins spent nuclear fuel has been developed and is presented in Volume I of this document. The recommendation constitutes an aggressive series of projects to construct andmore » operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. The overall processing and storage scheme is based on the ITA team`s proposed passivation and vault storage process. A dual purpose staging and vault storage facility provides an innovative feature which allows accelerated removal of fuel and sludge from the basins and minimizes programmatic risks beyond any of the originally proposed alternatives. The projects fit within a regulatory and National Environmental Policy Act (NEPA) overlay which mandates a two-phased approach to construction and operation of the needed facilities. The two-phase strategy packages and moves K Basins fuel and sludge to a newly constructed Staging and Storage Facility by the year 2000 where it is staged for processing. When an adjoining facility is constructed, the fuel is cycled through a stabilization process and returned to the Staging and Storage Facility for dry interim (40-year) storage. The estimated total expenditure for this Recommended Path Forward, including necessary new construction, operations, and deactivation of Project facilities through 2012, is approximately $1,150 million (unescalated).« less

Cermet-fueled reactors for advanced space applications

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

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

Process for making ceramic insulation

DOEpatents

Akash, Akash [Salt Lake City, UT; Balakrishnan, G Nair [Sandy, UT

2009-12-08

A method is provided for producing insulation materials and insulation for high temperature applications using novel castable and powder-based ceramics. The ceramic components produced using the proposed process offers (i) a fine porosity (from nano-to micro scale); (ii) a superior strength-to-weight ratio; and (iii) flexibility in designing multilayered features offering multifunctionality which will increase the service lifetime of insulation and refractory components used in the solid oxide fuel cell, direct carbon fuel cell, furnace, metal melting, glass, chemical, paper/pulp, automobile, industrial heating, coal, and power generation industries. Further, the ceramic components made using this method may have net-shape and/or net-size advantages with minimum post machining requirements.

78 FR 66858 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

...-2012-0246] RIN 3150-AJ20 Waste Confidence--Continued Storage of Spent Nuclear Fuel AGENCY: Nuclear... its generic determination on the environmental impacts of the continued storage of spent nuclear fuel... revising the generic determination of the environmental impacts of the continued storage of spent nuclear...

Technology readiness levels for advanced nuclear fuels and materials development

DOE PAGES

Carmack, W. J.; Braase, L. A.; Wigeland, R. A.; ...

2016-12-23

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. It was pioneered by the National Aeronautics and Space Administration (NASA) in the 1980s to develop and deploy new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications as well as the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is a critical technology needed for improving the performance and safety of currentmore » and advanced reactors, and ultimately closing the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management, communication and tracking tool. Furthermore, this article provides examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).« less

Technology readiness levels for advanced nuclear fuels and materials development

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

Carmack, W. J.; Braase, L. A.; Wigeland, R. A.

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. It was pioneered by the National Aeronautics and Space Administration (NASA) in the 1980s to develop and deploy new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications as well as the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is a critical technology needed for improving the performance and safety of currentmore » and advanced reactors, and ultimately closing the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management, communication and tracking tool. Furthermore, this article provides examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).« less

Exceptional power density and stability at intermediate temperatures in protonic ceramic fuel cells

NASA Astrophysics Data System (ADS)

Choi, Sihyuk; Kucharczyk, Chris J.; Liang, Yangang; Zhang, Xiaohang; Takeuchi, Ichiro; Ji, Ho-Il; Haile, Sossina M.

2018-03-01

Over the past several years, important strides have been made in demonstrating protonic ceramic fuel cells (PCFCs). Such fuel cells offer the potential of environmentally sustainable and cost-effective electric power generation. However, their power outputs have lagged behind predictions based on their high electrolyte conductivities. Here we overcome PCFC performance and stability challenges by employing a high-activity cathode, PrBa0.5Sr0.5Co1.5Fe0.5O5+δ (PBSCF), in combination with a chemically stable electrolyte, BaZr0.4Ce0.4Y0.1Yb0.1O3 (BZCYYb4411). We deposit a thin dense interlayer film of the cathode material onto the electrolyte surface to mitigate contact resistance, an approach which is made possible by the proton permeability of PBSCF. The peak power densities of the resulting fuel cells exceed 500 mW cm-2 at 500 °C, while also offering exceptional, long-term stability under CO2.

Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

DTIC Science & Technology

2008-01-20

critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation policy discouraged commercial nuclear fuel...have U.S. government officials. However, the case of Iran raises perhaps the most critical question in this decade for strengthening the nuclear...slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of uranium U3O8 to produce 1 lb of low

Ceramics breakthrough

NASA Astrophysics Data System (ADS)

Shim, Joon Hyung

2018-03-01

Protonic ceramic fuel cells (PCFCs) are a potential alternative to solid oxide fuel cells at temperatures below 600 °C, but they suffer from low power output and poor chemical stability. Now, PCFCs with a power density of over 500 mW per cm2 at 500 °C and long-term stability are demonstrated using optimized perovskite-based electrolyte and electrodes.

SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

DOEpatents

Schluderberg, D.C.; Ryon, J.W.

1962-05-01

A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

Locking support for nuclear fuel assemblies

DOEpatents

Ledin, Eric

1980-01-01

A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

Strengthening the nuclear nonproliferation regime : focus on the civilian nuclear fuel cycle.

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

Saltiel, David H.; Pregenzer, Arian Leigh

2005-04-01

Leaders around the world and across the ideological spectrum agree that the global nonproliferation regime is facing a serious test. The emergence of sophisticated terrorist networks, black markets in nuclear technology, and technological leaps associated with globalization have conspired to threaten one of the most successful examples of international cooperation in history. The rampant proliferation of nuclear weapons that was predicted at the start of the nuclear age has been largely held in check and the use of those weapons avoided. Nonetheless, with the thirty-fifth anniversary of the Treaty on the Nonproliferation of Nuclear Weapons (NPT), the threat of nuclearmore » proliferation seems more serious than ever. Although experts readily concede that there exist many pathways to proliferation, the threat posed by the misuse of the civilian nuclear fuel cycle has received considerable recent attention. While the connection between nuclear energy and nonproliferation has been a topic of discussion since the dawn of the nuclear age, world events have brought the issue to the forefront once again. United States President George W. Bush and International Atomic Energy Agency (IAEA) Director General Mohammad ElBaradei are among those who have highlighted proliferation risks associated with civilian nuclear power programs and called for revitalizing the nuclear nonproliferation regime to address new threats. From the possibility of diversion or theft of nuclear material or technology, to the use of national civilian programs as a cover for weapons programs - what some have called latent proliferation - the fuel cycle appears to many to represent a glaring proliferation vulnerability. Just as recognition of these risks is not new, neither is recognition of the many positive benefits of nuclear energy. In fact, a renewed interest in exploiting these benefits has increased the urgency of addressing the risks. Global energy demand is expected to at least double by the

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.

Mechanical and thermal characterization of a ceramic/glass composite seal for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Dev, Bodhayan; Walter, Mark E.; Arkenberg, Gene B.; Swartz, Scott L.

2014-01-01

Solid oxide fuel cells (SOFCs) require seals that can function in harsh, elevated temperature environments. Comprehensive characterization and understanding of seals is needed for commercially viable SOFCs. The present research focuses on a novel ceramic/glass composite seal that is produced by roller compaction or tape casting of glass and ceramic powders and an organic binder. Upon heat treatment, micro-voids and surface anomalies are formed. Increased heating and cooling rates during the heat treatment resulted in more and larger voids. The first goal of the current research is to suggest an appropriate heating and cooling rate to minimize the formation of microstructural defects. After identifying an appropriate cure cycle, seals were thermally cycled and then characterized with laser dilatometry, X-ray diffraction, and sonic resonance. From these experiments the crystalline phases, thermal expansion, and elastic properties were determined. Subsequently compression testing with an acoustic emission (AE) sensor and post-test microstructural analysis were used to identify the formation of damage. By fully understanding the characteristics of this ceramic/glass composite seal, next generation seals can be fabricated for improved performance.

The use of nuclear data in the field of nuclear fuel recycling

NASA Astrophysics Data System (ADS)

Martin, Julie-Fiona; Launay, Agnès; Grassi, Gabriele; Binet, Christophe; Lelandais, Jacques; Lecampion, Erick

2017-09-01

AREVA NC La Hague facility is the first step of the nuclear fuel recycling process implemented in France. The processing of the used fuel is governed by high standards of criticality-safety, and strong expectations on the quality of end-products. From the received used fuel assemblies, the plutonium and the uranium are extracted for further energy production purposes within the years following the reprocessing. Furthermore, the ultimate waste - fission products and minor actinides on the one hand, and hulls and end-pieces on the other hand - is adequately packaged for long term disposal. The used fuel is therefore separated into very different materials, and time scales which come into account may be longer than in some other nuclear fields of activity. Given the variety of the handled nuclear materials, as well as the time scales at stake, the importance given to some radionuclides, and hence to the associated nuclear data, can also be specific to the AREVA NC La Hague plant. A study has thus been led to identify a list of the most important radionuclides for the AREVA NC La Hague plant applications, relying on the running constraints of the facility, and the end-products expectations. The activities at the AREVA NC La Hague plant are presented, and the methodology to extract the most important radionuclides for the reprocessing process is detailed.

Spent Nuclear Fuel Alternative Technology Decision Analysis

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

Shedrow, C.B.

1999-11-29

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies andmore » ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.« less

Unirradiated testing of the demonstration-scale ceramic waste form at ANL-West

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

Goff, K.M.; Simpson, M.F.; Bateman, K.J.

1997-12-01

The ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel for disposal. The alkali, alkaline earth, halide, and rare earth fission products are stabilized in zeolite, which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The transuranics, including plutonium, are also stabilized in this high-level waste. Most of the laboratory-scale development work is performed in the Chemical Technology Division of ANL in Illinois. At ANL-West in Idaho, this technology is being demonstrated on an engineering scalemore » before implementation with irradiated materials in a remote environment.« less

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

DOE PAGES

Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; ...

2015-09-10

Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Biokinetics of nuclear fuel compounds and biological effects of nonuniform radiation.

PubMed Central

Lang, S; Servomaa, K; Kosma, V M; Rytömaa, T

1995-01-01

Environmental releases of insoluble nuclear fuel compounds may occur at nuclear power plants during normal operation, after nuclear power plant accidents, and as a consequence of nuclear weapons testing. For example, the Chernobyl fallout contained extensive amounts of pulverized nuclear fuel composed of uranium and its nonvolatile fission products. The effects of these highly radioactive particles, also called hot particles, on humans are not well known due to lack of reliable data on the extent of the exposure. However, the biokinetics and biological effects of nuclear fuel compounds have been investigated in a number of experimental studies using various cellular systems and laboratory animals. In this article, we review the biokinetic properties and effects of insoluble nuclear fuel compounds, with special reference to UO2, PuO2, and nonvolatile, long-lived beta-emitters Zr, Nb, Ru, and Ce. First, the data on hot particles, including sources, dosimetry, and human exposure are discussed. Second, the biokinetics of insoluble nuclear fuel compounds in the gastrointestinal tract and respiratory tract are reviewed. Finally, short- and long-term biological effects of nonuniform alpha- and beta-irradiation on the gastrointestinal tract, lungs, and skin are discussed. Images p920-a Figure 1. PMID:8529589

Fuel handling apparatus for a nuclear reactor

DOEpatents

Hawke, Basil C.

1987-01-01

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

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)

Fractal Model of Fission Product Release in Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stankunas, Gediminas

2012-09-01

A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald-Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald-Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald-Letnikov derivative parameter with fuel burn-up is established.

Method and means of packaging nuclear fuel rods for handling

DOEpatents

Adam, Milton F.

1979-01-01

Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

Fire resistant nuclear fuel cask

DOEpatents

Heckman, Richard C.; Moss, Marvin

1979-01-01

The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

A cermet fuel reactor for nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Kruger, Gordon

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

Nuclear Waste Disposal and Strategies for Predicting Long-Term Performance of Material

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

Wicks, G G

2001-03-28

Ceramics have been an important part of the nuclear community for many years. On December 2, 1942, an historic event occurred under the West Stands of Stagg Field, at the University of Chicago. Man initiated his first self-sustaining nuclear chain reaction and controlled it. The impact of this event on civilization is considered by many as monumental and compared by some to other significant events in history, such as the invention of the steam engine and the manufacturing of the first automobile. Making this event possible and the successful operation of this first man-made nuclear reactor, was the use ofmore » forty tons of UO2. The use of natural or enriched UO2 is still used today as a nuclear fuel in many nuclear power plants operating world-wide. Other ceramic materials, such as 238Pu, are used for other important purposes, such as ceramic fuels for space exploration to provide electrical power to operate instruments on board spacecrafts. Radioisotopic Thermoelectric Generators (RTGs) are used to supply electrical power and consist of a nuclear heat source and converter to transform heat energy from radioactive decay into electrical power, thus providing reliable and relatively uniform power over the very long lifetime of a mission. These sources have been used in the Galileo spacecraft orbiting Jupiter and for scientific investigations of Saturn with the Cassini spacecraft. Still another very important series of applications using the unique properties of ceramics in the nuclear field, are as immobilization matrices for management of some of the most hazardous wastes known to man. For example, in long-term management of radioactive and hazardous wastes, glass matrices are currently in production immobilizing high-level radioactive materials, and cementious forms have also been produced to incorporate low level wastes. Also, as part of nuclear disarmament activities, assemblages of crystalline phases are being developed for immobilizing weapons grade

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.

Means for supporting fuel elements in a nuclear reactor

DOEpatents

Andrews, Harry N.; Keller, Herbert W.

1980-01-01

A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

Ceramic waste form production and development at ANL-West.

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

Battisti, T. J.; Goff, K. M.; Bateman, K. J.

2002-08-21

Argonne National Laboratory has developed a method to stabilize spent electrolyte salt discarded from electrorefiners (ER) used to treat spent nuclear fuel. The salt is stabilized in a ceramic using a pressureless consolidation technique. The starting material is zeolite 4A which is used as the host for the fission product and actinide rich salt. Glass frit is added to the salt loaded zeolite before processing to act as a binder. The zeolite 4A is converted to sodalite during processing via pressureless consolidation. This process differs from one used in the past that employed a hot isostatic press. Ceramic is createdmore » at 925 C and atmospheric pressure instead of the high pressures used in hot isostatic pressing. Process flow sheets, off-gas test results, processing equipment, and leech test results are presented.« less

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

2017-09-01

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Spent nuclear fuel discharges from US reactors 1992

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

Not Available

1994-05-05

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactivemore » Waste Management.« less

Financing Strategies For A Nuclear Fuel Cycle Facility

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

David Shropshire; Sharon Chandler

2006-07-01

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest lifecycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due tomore » government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

Fuel element concept for long life high power nuclear reactors

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.; Rom, F. E.

1969-01-01

Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

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

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« less

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

Nuclear fuel alloys or mixtures and method of making thereof

DOEpatents

Mariani, Robert Dominick; Porter, Douglas Lloyd

2016-04-05

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

Application of X-ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

DOE PAGES

Silva, Chinthaka M.; Snead, Lance Lewis; Hunn, John D.; ...

2015-08-03

X-ray microcomputed tomography (µCT) was applied in characterizing the internal structures of a number of irradiated materials, including carbon-carbon fibre composites, nuclear-grade graphite and tristructural isotropic-coated fuel particles. Local cracks in carbon-carbon fibre composites associated with their synthesis process were observed with µCT without any destructive sample preparation. Pore analysis of graphite samples was performed quantitatively, and qualitative analysis of pore distribution was accomplished. It was also shown that high-resolution µCT can be used to probe internal layer defects of tristructural isotropic-coated fuel particles to elucidate the resulting high release of radioisotopes. Layer defects of sizes ranging from 1 tomore » 5 µm and up could be isolated by to-mography. As an added advantage, µCT could also be used to identify regions with high densities of radioisotopes to deter-mine the proper plane and orientation of particle mounting for further analytical characterization, such as materialographic sectioning followed by optical and electron microscopy. Lastly, in fully ceramic matrix fuel forms, despite the highly absorbing matrix, characterization of tristructural isotropic-coated particles embedded in a silicon carbide matrix was accomplished usingµCT and related advanced image analysis techniques.« less

Integration Science and Technology of Advanced Ceramics for Energy and Environmental Applications

NASA Technical Reports Server (NTRS)

Singh, M.

2012-01-01

The discovery of new and innovative materials has been known to culminate in major turning points in human history. The transformative impact and functional manifestation of new materials have been demonstrated in every historical era by their integration into new products, systems, assemblies, and devices. In modern times, the integration of new materials into usable products has a special relevance for the technological development and economic competitiveness of industrial societies. Advanced ceramic technologies dramatically impact the energy and environmental landscape due to potential wide scale applications in all aspects of energy production, storage, distribution, conservation, and efficiency. Examples include gas turbine propulsion systems, fuel cells, thermoelectrics, photovoltaics, distribution and transmission systems based on superconductors, nuclear power generation, and waste disposal. Robust ceramic integration technologies enable hierarchical design and manufacturing of intricate ceramic components starting with geometrically simpler units that are subsequently joined to themselves and/or to metals to create components with progressively higher levels of complexity and functionality. However, for the development of robust and reliable integrated systems with optimum performance under different operating conditions, the detailed understanding of various thermochemical and thermomechanical factors is critical. Different approaches are required for the integration of ceramic-metal and ceramic-ceramic systems across length scales (macro to nano). In this presentation, a few examples of integration of ceramic to metals and ceramic to ceramic systems will be presented. Various challenges and opportunities in design, fabrication, and testing of integrated similar (ceramic-ceramic) and dissimilar (ceramic-metal) material systems will be discussed. Potential opportunities and need for the development of innovative design philosophies, approaches, and

Trace and surface analysis of ceramic layers of solid oxide fuel cells by mass spectrometry.

PubMed

Becker, J S; Breuer, U; Westheide, J; Saprykin, A I; Holzbrecher, H; Nickel, H; Dietze, H J

1996-06-01

For the trace analysis of impurities in thick ceramic layers of a solid oxide fuel cell (SOFC) sensitive solid-state mass spectrometric methods, such as laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and radiofrequency glow discharge mass spectrometry (rf-GDMS) have been developed and used. In order to quantify the analytical results of LA-ICP-MS, the relative sensitivity coefficients of elements in a La(0.6)Sr(0.35)MnO(3) matrix have been determined using synthetic standards. Secondary ion mass spectrometry (SIMS) - as a surface analytical method - has been used to characterize the element distribution and diffusion profiles of matrix elements on the interface of a perovskite/Y-stabilized ZrO(2) layer. The application of different mass spectrometric methods for process control in the preparation of ceramic layers for the SOFC is described.

Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

DTIC Science & Technology

2008-09-03

Spent nuclear fuel disposal has remained the most critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation...inalienable right and by and large, neither have U.S. government officials. However, the case of Iran raises perhaps the most critical question in...the enrichment process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires

Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar

2012-01-01

A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

Pyroprocess for processing spent nuclear fuel

DOEpatents

Miller, William E.; Tomczuk, Zygmunt

2002-01-01

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

Plutonium: Advancing our Understanding to Support Sustainable Nuclear Fuel Cycles

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

Lines, Amanda M.; Adami, Susan R.; Casella, Amanda

With Global energy needs increasing, real energy solutions to meet demands now, are needed. Fossil fuels are not an ideal candidate to meet these needs because of their negative impact on the environment. Renewables such as wind and solar have huge potential, but still need major technological advancements (particularly in the area of battery storage) before they can effectively meet growing world needs. The best option for meeting large energy needs without a large carbon footprint is nuclear energy. Of course, nuclear energy can face a fair amount of opposition and concern. However, through modern engineering and science many ofmore » these concerns can now be addressed. Many safety concerns can be met by engineering advancements, but perhaps the biggest area of concern is what to do with the used nuclear fuel after it is removed from the reactor. Currently the United States (and several other countries) utilize an open fuel cycle, meaning fuel is only used once and then discarded. It should be noted that fuel coming out of a reactor has utilized approximately 1% of the total energy that could be produced by the uranium in the fuel rod. The answer here is to close the fuel cycle and recycle the nuclear materials. By reprocessing used nuclear fuel, all the U can be repurposed without requiring disposal. The various fission products can be removed and either discarded (hugely reduced waste volume) or more reasonably, utilized in specialty reactors to make more energy or needed research/medical isotopes. While reprocessing technology is currently advanced enough to meet energy needs, completing research to improve and better understand these techniques is still needed. Better understanding behavior of fission products is one area of important research. Despite it being discovered over 75 years ago, plutonium is still an exciting element to study because of the complex solution chemistry it exhibits. In aqueous solutions Pu can exist simultaneously in multiple

Classical Molecular Dynamics Simulation of Nuclear Fuel

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

Devanathan, Ram; Krack, Matthias; Bertolus, Marjorie

2015-10-10

Molecular dynamics simulation is well suited to study primary damage production by irradiation, defect interactions with fission gas atoms, gas bubble nucleation, grain boundary effects on defect and gas bubble evolution in nuclear fuel, and the resulting changes in thermo-mechanical properties. In these simulations, the forces on the ions are dictated by interaction potentials generated by fitting properties of interest to experimental data. The results obtained from the present generation of potentials are qualitatively similar, but quantitatively different. There is a need to refine existing potentials to provide a better representation of the performance of polycrystalline fuel under a varietymore » of operating conditions, and to develop models that are equipped to handle deviations from stoichiometry. In addition to providing insights into fundamental mechanisms governing the behaviour of nuclear fuel, MD simulations can also provide parameters that can be used as inputs for mesoscale models.« less

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

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

N /A

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign andmore » domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.« less

Nuclear fuel microsphere gamma analyzer

DOEpatents

Valentine, Kenneth H.; Long, Jr., Ernest L.; Willey, Melvin G.

1977-01-01

A gamma analyzer system is provided for the analysis of nuclear fuel microspheres and other radioactive particles. The system consists of an analysis turntable with means for loading, in sequence, a plurality of stations within the turntable; a gamma ray detector for determining the spectrum of a sample in one section; means for analyzing the spectrum; and a receiver turntable to collect the analyzed material in stations according to the spectrum analysis. Accordingly, particles may be sorted according to their quality; e.g., fuel particles with fractured coatings may be separated from those that are not fractured, or according to other properties.

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.

Proliferation resistance assessment of various methods of spent nuclear fuel storage and disposal

NASA Astrophysics Data System (ADS)

Kollar, Lenka

Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for

Nuclear proliferation-resistance and safeguards for future nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Kuno, Y.; Inoue, N.; Senzaki, M.

2009-03-01

Corresponding to the world nuclear security concerns, future nuclear fuel cycle (NFC) should have high proliferation-resistance (PR) and physical protection (PP), while promotion of the peaceful use of the nuclear energy must not be inhibited. In order to accomplish nuclear non-proliferation from NFC, a few models of the well-PR systems should be developed so that international community can recognize them as worldwide norms. To find a good balance of 'safeguard-ability (so-called extrinsic measure or institutional barrier)' and 'impede-ability (intrinsic feature or technical barrier)' will come to be essential for NFC designers to optimize civilian nuclear technology with nuclear non-proliferation, although the advanced safeguards with high detectability can still play a dominant role for PR in the states complying with full institutional controls. Accomplishment of such goal in a good economic efficiency is a future key challenge.

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

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

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

Metals and Ceramics Division progress report for period ending December 31, 1992

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

Craig, D.F.; Weir, J.R. Jr.

1993-04-01

This report provides a brief overview of the activities and accomplishments of the division, whose purpose is to provide technical support, primarily in the area of high-temperature materials, for the various technologies being developed by US DOE. Activities range from basic research to industrial research and technology transfer. The division (and the report) is divided into the following: Engineering materials, high-temperature materials, materials science, ceramics, nuclear fuel materials, program activities, collaborative research facilities and technology transfer, and educational programs.

Comparing the short and long term stability of biodegradable, ceramic and cation exchange membranes in microbial fuel cells.

PubMed

Winfield, Jonathan; Chambers, Lily D; Rossiter, Jonathan; Ieropoulos, Ioannis

2013-11-01

The long and short-term stability of two porous dependent ion exchange materials; starch-based compostable bags (BioBag) and ceramic, were compared to commercially available cation exchange membrane (CEM) in microbial fuel cells. Using bi-directional polarisation methods, CEM exhibited power overshoot during the forward sweep followed by significant power decline over the reverse sweep (38%). The porous membranes displayed no power overshoot with comparably smaller drops in power during the reverse sweep (ceramic 8%, BioBag 5.5%). The total internal resistance at maximum power increased by 64% for CEM compared to 4% (ceramic) and 6% (BioBag). Under fixed external resistive loads, CEM exhibited steeper pH reductions than the porous membranes. Despite its limited lifetime, the BioBag proved an efficient material for a stable microbial environment until failing after 8 months, due to natural degradation. These findings highlight porous separators as ideal candidates for advancing MFC technology in terms of cost and operation stability. Copyright © 2013 Elsevier Ltd. All rights reserved.

Air Shipment of Spent Nuclear Fuel from Romania to Russia

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

Igor Bolshinsky; Ken Allen; Lucian Biro

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities formore » shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.« less

Ceramic Borehole Seals for Nuclear Waste Disposal Applications

NASA Astrophysics Data System (ADS)

Lowry, B.; Coates, K.; Wohletz, K.; Dunn, S.; Patera, E.; Duguid, A.; Arnold, B.; Zyvoloski, G.; Groven, L.; Kuramyssova, K.

2015-12-01

Sealing plugs are critical features of the deep borehole system design. They serve as structural platforms to bear the weight of the backfill column, and as seals through their low fluid permeability and bond to the borehole or casing wall. High hydrostatic and lithostatic pressures, high mineral content water, and elevated temperature due to the waste packages and geothermal gradient challenge the long term performance of seal materials. Deep borehole nuclear waste disposal faces the added requirement of assuring performance for thousands of years in large boreholes, requiring very long term chemical and physical stability. A high performance plug system is being developed which capitalizes on the energy of solid phase reactions to form a ceramic plug in-situ. Thermites are a family of self-oxidized metal/oxide reactions with very high energy content and the ability to react under water. When combined with engineered additives the product exhibits attractive structural, sealing, and corrosion properties. In the initial phase of this research, exploratory and scaled tests demonstrated formulations that achieved controlled, fine grained, homogeneous, net shape plugs composed predominantly of ceramic material. Laboratory experiments produced plug cores with confined fluid permeability as low as 100 mDarcy, compressive strength as high as 70 MPa (three times the strength of conventional well cement), with the inherent corrosion resistance and service temperature of ceramic matrices. Numerical thermal and thermal/structural analyses predicted the in-situ thermal performance of the reacted plugs, showing that they cooled to ambient temperature (and design strength) within 24 to 48 hours. The current development effort is refining the reactant formulations to achieve desired performance characteristics, developing the system design and emplacement processes to be compatible with conventional well service practices, and understanding the thermal, fluid, and structural

Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Panda, Binayak; Shah, Sandeep

2005-01-01

Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

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.

Systems for the Intermodal Routing of Spent Nuclear Fuel

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

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the

Hazards of the Nuclear Fuel Cycle

ERIC Educational Resources Information Center

Holdren, John P.

1974-01-01

Outlines the stages of the nuclear fuel cycle where routine radiation releases occur and where nonroutine releases could occur. Examines the impact of these occurrences and emphasizes the regulations, practices, and technologies that prevail in the United States. (Author/GS)

Asymptotic Expansion Homogenization for Multiscale Nuclear Fuel Analysis

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

Hales, J. D.; Tonks, M. R.; Chockalingam, K.

2015-03-01

Engineering scale nuclear fuel performance simulations can benefit by utilizing high-fidelity models running at a lower length scale. Lower length-scale models provide a detailed view of the material behavior that is used to determine the average material response at the macroscale. These lower length-scale calculations may provide insight into material behavior where experimental data is sparse or nonexistent. This multiscale approach is especially useful in the nuclear field, since irradiation experiments are difficult and expensive to conduct. The lower length-scale models complement the experiments by influencing the types of experiments required and by reducing the total number of experiments needed.more » This multiscale modeling approach is a central motivation in the development of the BISON-MARMOT fuel performance codes at Idaho National Laboratory. These codes seek to provide more accurate and predictive solutions for nuclear fuel behavior. One critical aspect of multiscale modeling is the ability to extract the relevant information from the lower length-scale sim- ulations. One approach, the asymptotic expansion homogenization (AEH) technique, has proven to be an effective method for determining homogenized material parameters. The AEH technique prescribes a system of equations to solve at the microscale that are used to compute homogenized material constants for use at the engineering scale. In this work, we employ AEH to explore the effect of evolving microstructural thermal conductivity and elastic constants on nuclear fuel performance. We show that the AEH approach fits cleanly into the BISON and MARMOT codes and provides a natural, multidimensional homogenization capability.« less

Freeze Tape Casting of Functionally Graded Porous Ceramics

NASA Technical Reports Server (NTRS)

Sofie, Stephen W.

2007-01-01

Freeze tape casting is a means of making preforms of ceramic sheets that, upon subsequent completion of fabrication processing, can have anisotropic and/or functionally graded properties that notably include aligned and graded porosity. Freeze tape casting was developed to enable optimization of the microstructures of porous ceramic components for use as solid oxide electrodes in fuel cells: Through alignment and grading of pores, one can tailor surface areas and diffusion channels for flows of gas and liquid species involved in fuel-cell reactions. Freeze tape casting offers similar benefits for fabrication of optimally porous ceramics for use as catalysts, gas sensors, and filters.

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

DOEpatents

Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

2013-11-05

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

DOEpatents

Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

2014-06-10

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2013 CFR

2013-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2014 CFR

2014-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

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.

Rapid methods for radionuclide contaminant transport in nuclear fuel cycle simulation

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

Huff, Kathryn

Here, nuclear fuel cycle and nuclear waste disposal decisions are technologically coupled. However, current nuclear fuel cycle simulators lack dynamic repository performance analysis due to the computational burden of high-fidelity hydrolgic contaminant transport models. The Cyder disposal environment and repository module was developed to fill this gap. It implements medium-fidelity hydrologic radionuclide transport models to support assessment appropriate for fuel cycle simulation in the Cyclus fuel cycle simulator. Rapid modeling of hundreds of discrete waste packages in a geologic environment is enabled within this module by a suite of four closed form models for advective, dispersive, coupled, and idealized con-more » taminant transport: a Degradation Rate model, a Mixed Cell model, a Lumped Parameter model, and a 1-D Permeable Porous Medium model. A summary of the Cyder module, its timestepping algorithm, and the mathematical models implemented within it are presented. Additionally, parametric demonstrations simulations performed with Cyder are presented and shown to demonstrate functional agreement with parametric simulations conducted in a standalone hydrologic transport model, the Clay Generic Disposal System Model developed by the Used Fuel Disposition Campaign Department of Energy Office of Nuclear Energy.« less

Rapid methods for radionuclide contaminant transport in nuclear fuel cycle simulation

DOE PAGES

Huff, Kathryn

2017-08-01

Here, nuclear fuel cycle and nuclear waste disposal decisions are technologically coupled. However, current nuclear fuel cycle simulators lack dynamic repository performance analysis due to the computational burden of high-fidelity hydrolgic contaminant transport models. The Cyder disposal environment and repository module was developed to fill this gap. It implements medium-fidelity hydrologic radionuclide transport models to support assessment appropriate for fuel cycle simulation in the Cyclus fuel cycle simulator. Rapid modeling of hundreds of discrete waste packages in a geologic environment is enabled within this module by a suite of four closed form models for advective, dispersive, coupled, and idealized con-more » taminant transport: a Degradation Rate model, a Mixed Cell model, a Lumped Parameter model, and a 1-D Permeable Porous Medium model. A summary of the Cyder module, its timestepping algorithm, and the mathematical models implemented within it are presented. Additionally, parametric demonstrations simulations performed with Cyder are presented and shown to demonstrate functional agreement with parametric simulations conducted in a standalone hydrologic transport model, the Clay Generic Disposal System Model developed by the Used Fuel Disposition Campaign Department of Energy Office of Nuclear Energy.« less

Method of preparing thin porous sheets of ceramic material

DOEpatents

Swarr, Thomas E.; Nickols, Richard C.; Krasij, Myron

1987-03-24

A method of forming thin porous sheets of ceramic material for use as electrodes or other components in a molten carbonate fuel cell is disclosed. The method involves spray drying a slurry of fine ceramic particles in liquid carrier to produce generally spherical agglomerates of high porosity and a rough surface texture. The ceramic particles may include the electrode catalyst and the agglomerates can be calcined to improve mechanical strength. After slurrying with suitable volatile material and binder tape casting is used to form sheets that are sufficiently strong for further processing and handling in the assembly of a high temperature fuel cell.

Method of preparing thin porous sheets of ceramic material

DOEpatents

Swarr, T.E.; Nickols, R.C.; Krasij, M.

1984-05-23

A method of forming thin porous sheets of ceramic material for use as electrodes or other components in a molten carbonate fuel cell is disclosed. The method involves spray drying a slurry of fine ceramic particles in liquid carrier to produce generally spherical agglomerates of high porosity and a rough surface texture. The ceramic particles may include the electrode catalyst and the agglomerates can be calcined to improve mechanical strength. After slurrying with suitable volatile material and binder tape casting is used to form sheets that are sufficiently strong for further processing and handling in the assembly of a high temperature fuel cell.

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

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

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

Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

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

Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.

2011-11-14

For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may bemore » constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.« less

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

DOEpatents

Flint, O.

1961-01-10

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Development of advanced test methods for the improvement of production standards for ceramic powders used in solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Ward, Brian

Solid oxide fuel cells (SOFCs) are energy conversion devices that use ceramic powders as a precursor material for their electrodes. Presently, powder manufacturers are encountering complications producing consistent precursor powders. Through various thermal, chemical and physical tests, such as DSC and XRD, a preliminary production standard will be developed.

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

On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

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

Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F., E-mail: Tsibulskiy-VF@nrcki.ru

2015-12-15

Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

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

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrixmore » composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing« less

Ceramic high pressure gas path seal

NASA Technical Reports Server (NTRS)

Liotta, G. C.

1987-01-01

Stage 1 ceramic shrouds (high pressure turbine gas path seal) were developed for the GE T700 turbine helicopter engine under the Army/NASA Contract NAS3-23174. This contract successfully proved the viability and benefits of a Stage 1 ceramic shroud for production application. Stage 1 ceramic shrouds were proven by extensive component and engine testing. This Stage 1 ceramic shroud, plasma sprayed ceramic (ZrOs-BY2O3) and bond coating (NiCrAlY) onto a cast metal backing, offers significant engine performance improvement. Due to the ceramic coating, the amount of cooling air required is reduced 20% resulting in a 0.5% increase in horsepower and a 0.3% decrease in specific fuel consumption. This is accomplished with a component which is lower in cost than the current production shroud. Stage 1 ceramic shrouds will be introduced into field service in late 1987.

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

Compact Fuel Element Environment Test

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

2012-01-01

Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

Temperature measuring analysis of the nuclear reactor fuel assembly

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

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

Signatures of Extended Storage of Used Nuclear Fuel in Casks

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

Rauch, Eric Benton

2016-09-28

As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Controlmore » Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.« less

Study and program plan for improved heavy duty gas turbine engine ceramic component development

NASA Technical Reports Server (NTRS)

Helms, H. E.

1977-01-01

Fuel economy in a commercially viable gas turbine engine was demonstrated through use of ceramic materials. Study results show that increased turbine inlet and generator inlet temperatures, through the use of ceramic materials, contribute the greatest amount to achieving fuel economy goals. Improved component efficiencies show significant additional gains in fuel economy.

Nuclear rocket using indigenous Martian fuel NIMF

NASA Technical Reports Server (NTRS)

Zubrin, Robert

1991-01-01

In the 1960's, Nuclear Thermal Rocket (NTR) engines were developed and ground tested capable of yielding isp of up to 900 s at thrusts up to 250 klb. Numerous trade studies have shown that such traditional hydrogen fueled NTR engines can reduce the inertial mass low earth orbit (IMLEO) of lunar missions by 35 percent and Mars missions by 50 to 65 percent. The same personnel and facilities used to revive the hydrogen NTR can also be used to develop NTR engines capable of using indigenous Martian volatiles as propellant. By putting this capacity of the NTR to work in a Mars descent/acent vehicle, the Nuclear rocket using Indigenous Martian Fuel (NIMF) can greatly reduce the IMLEO of a manned Mars mission, while giving the mission unlimited planetwide mobility.

Electron Correlation and Tranport Properties in Nuclear Fuel Materials

NASA Astrophysics Data System (ADS)

Yin, Quan; Haule, Kristjan; Kotliar, Gabriel; Savrasov, Sergey; Pickett, Warren

2011-03-01

Using first principle LDA+DMFT method, we conduct a systematic study on the correlated electronic structures and transport properties of select actinide carbides, nitrides, and oxides, many of which are nuclear fuel materials. Our results capture the metal--insulator Mott transition within the studied systems, and the appearance of the Zhang-Rice state in uranium dioxide. More importantly, by understanding the physics underlying their transport properties, we suggest ways to improve the efficiency of currently used fuels. This work is supported by the DOE Nuclear Energy University Program, contract No. 00088708.

Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

2018-03-01

Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

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

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

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

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information

Casting Of Multilayer Ceramic Tapes

NASA Technical Reports Server (NTRS)

Collins, Earl R., Jr.

1991-01-01

Procedure for casting thin, multilayer ceramic membranes, commonly called tapes, involves centrifugal casting at accelerations of 1,800 to 2,000 times normal gravitational acceleration. Layers of tape cast one at a time on top of any previous layer or layers. Each layer cast from slurry of ground ceramic suspended in mixture of solvents, binders, and other components. Used in capacitors, fuel cells, and electrolytic separation of oxygen from air.

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

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

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

Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru

2015-12-15

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2012 CFR

2012-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the...

Evaluation of conventional power systems. [emphasizing fossil fuels and nuclear energy

NASA Technical Reports Server (NTRS)

Smith, K. R.; Weyant, J.; Holdren, J. P.

1975-01-01

The technical, economic, and environmental characteristics of (thermal, nonsolar) electric power plants are reviewed. The fuel cycle, from extraction of new fuel to final waste management, is included. Emphasis is placed on the fossil fuel and nuclear technologies.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

Platinized Graphene/ceramics Nano-sandwiched Architectures and Electrodes with Outstanding Performance for PEM Fuel Cells

PubMed Central

Chen, Xu; He, Daping; Wu, Hui; Zhao, Xiaofeng; Zhang, Jian; Cheng, Kun; Wu, Peng; Mu, Shichun

2015-01-01

For the first time a novel oxygen reduction catalyst with a 3D platinized graphene/nano-ceramic sandwiched architecture is successfully prepared by an unusual method. Herein the specific gravity of graphene nanosheets (GNS) is tailored by platinizing graphene in advance to shorten the difference in the specific gravity between carbon and SiC materials, and then nano-SiC is well intercalated into GNS interlayers. This nano-architecture with highly dispersed Pt nanoparticles exhibits a very high oxygen reduction reaction (ORR) activity and polymer electrolyte membrane (PEM) fuel cell performance. The mass activity of half cells is 1.6 times of that of the GNS supported Pt, and 2.4 times that of the commercial Pt/C catalyst, respectively. Moreover, after an accelerated stress test our catalyst shows a predominantly electrochemical stability compared with benchmarks. Further fuel cell tests show a maximum power density as high as 747 mW/cm2 at low Pt loading, which is more than 2 times higher than that of fuel cells with the pristine graphene electrode. PMID:26538366

Platinized Graphene/ceramics Nano-sandwiched Architectures and Electrodes with Outstanding Performance for PEM Fuel Cells.

PubMed

Chen, Xu; He, Daping; Wu, Hui; Zhao, Xiaofeng; Zhang, Jian; Cheng, Kun; Wu, Peng; Mu, Shichun

2015-11-05

For the first time a novel oxygen reduction catalyst with a 3D platinized graphene/nano-ceramic sandwiched architecture is successfully prepared by an unusual method. Herein the specific gravity of graphene nanosheets (GNS) is tailored by platinizing graphene in advance to shorten the difference in the specific gravity between carbon and SiC materials, and then nano-SiC is well intercalated into GNS interlayers. This nano-architecture with highly dispersed Pt nanoparticles exhibits a very high oxygen reduction reaction (ORR) activity and polymer electrolyte membrane (PEM) fuel cell performance. The mass activity of half cells is 1.6 times of that of the GNS supported Pt, and 2.4 times that of the commercial Pt/C catalyst, respectively. Moreover, after an accelerated stress test our catalyst shows a predominantly electrochemical stability compared with benchmarks. Further fuel cell tests show a maximum power density as high as 747 mW/cm(2) at low Pt loading, which is more than 2 times higher than that of fuel cells with the pristine graphene electrode.

Platinized Graphene/ceramics Nano-sandwiched Architectures and Electrodes with Outstanding Performance for PEM Fuel Cells

NASA Astrophysics Data System (ADS)

Chen, Xu; He, Daping; Wu, Hui; Zhao, Xiaofeng; Zhang, Jian; Cheng, Kun; Wu, Peng; Mu, Shichun

2015-11-01

For the first time a novel oxygen reduction catalyst with a 3D platinized graphene/nano-ceramic sandwiched architecture is successfully prepared by an unusual method. Herein the specific gravity of graphene nanosheets (GNS) is tailored by platinizing graphene in advance to shorten the difference in the specific gravity between carbon and SiC materials, and then nano-SiC is well intercalated into GNS interlayers. This nano-architecture with highly dispersed Pt nanoparticles exhibits a very high oxygen reduction reaction (ORR) activity and polymer electrolyte membrane (PEM) fuel cell performance. The mass activity of half cells is 1.6 times of that of the GNS supported Pt, and 2.4 times that of the commercial Pt/C catalyst, respectively. Moreover, after an accelerated stress test our catalyst shows a predominantly electrochemical stability compared with benchmarks. Further fuel cell tests show a maximum power density as high as 747 mW/cm2 at low Pt loading, which is more than 2 times higher than that of fuel cells with the pristine graphene electrode.

Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

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.

Benchmarking the expected stack manufacturing cost of next generation, intermediate-temperature protonic ceramic fuel cells with solid oxide fuel cell technology

NASA Astrophysics Data System (ADS)

Dubois, Alexis; Ricote, Sandrine; Braun, Robert J.

2017-11-01

Recent progress in the performance of intermediate temperature (500-600 °C) protonic ceramic fuel cells (PCFCs) has demonstrated both fuel flexibility and increasing power density that approach commercial application requirements. These developments may eventually position the technology as a viable alternative to solid oxide fuel cells (SOFCs) and molten carbonate fuel cells (MCFCs). The PCFCs investigated in this work are based on a BaZr0.8Y0.2O3-δ (BZY20) thin electrolyte supported by BZY20/Ni porous anodes, and a triple conducting cathode material comprised of BaCo0.4Fe0.4Zr0.1Y0.1O3-δ (BCFZY0.1). These cells are prepared using a low-cost solid-state reactive sintering (SSRS) process, and are capable of power densities of 0.156 W cm-2 at 500 °C operating directly from methane fuel. We develop a manufacturing cost model to estimate the Nth generation production costs of PCFC stack technology using high volume manufacturing processes and compare them to the state-of-the-art in SOFC technology. The low-cost cell manufacturing enabled by the SSRS technique compensates for the lower PCFC power density and the trade-off between operating temperature and efficiency enables the use of lower-cost stainless steel materials. PCFC stack production cost estimates are found to be as much as 27-37% lower at 550 °C than SOFCs operating at 800 °C.

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

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

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

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

Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Sooby, Elizabeth; Adams, Marvin; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Phongikaroon, Supathorn; Pogue, Nathaniel; Sattarov, Akhdiyor; Simpson, Michael; Tripathy, Prabhat; Tsevkov, Pavel

2013-04-01

The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

Nuclear reactor fuel element having improved heat transfer

DOEpatents

Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

1982-03-03

A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

Space Exploration Initiative Fuels, Materials and Related Nuclear Propulsion Technologies Panel

NASA Technical Reports Server (NTRS)

Bhattacharyya, S. K.; Olsen, C.; Cooper, R.; Matthews, R. B.; Walter, C.; Titran, R. J.

1993-01-01

This report was prepared by members of the Fuels, Materials and Related Technologies Panel, with assistance from a number of industry observers as well as laboratory colleagues of the panel members. It represents a consensus view of the panel members. This report was not subjected to a thorough review by DOE, NASA or DoD, and the opinions expressed should not be construed to represent the official position of these organizations, individually or jointly. Topics addressed include: requirement for fuels and materials development for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP); overview of proposed concepts; fuels technology development plan; materials technology development plan; other reactor technology development; and fuels and materials requirements for advanced propulsion concepts.

Preparation of nuclear fuel spheres by flotation-internal gelation

DOEpatents

Haas, Paul A.; Fowler, Victor L.; Lloyd, Milton H.

1987-01-01

A simplified internal gelation process for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85.degree. C.) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column.

Preparation of nuclear fuel spheres by flotation-internal gelation

DOEpatents

Haas, P.A.; Fowler, V.L.; Lloyd, M.H.

1984-12-21

A simplified internal gelation process is claimed for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85/sup 0/C) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column. 3 figs.

76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... processes are more akin to fuel cycle processes. This framework was established in the 1970's to license the... nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and deploying fast... Accounting;'' and a Nuclear Energy Institute white [[Page 34009

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)

Human capital needs - teaching, training and coordination for nuclear fuel cycle

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

Retegan, T.; Ekberg, C.; John, J.

Human capital is the accumulation of competencies, knowledge, social and creativity skills and personality attributes, which are necessary to perform work so as to produce economic value. In the frame of the nuclear fuel cycle, this is of paramount importance that the right human capital exists and in Europe this is fostered by a series of integrated or directed projects. The teaching, training and coordination will be discussed in the frame of University curricula with examples from several programs, like e.g. the Master of Nuclear Engineering at Chalmers University, Sweden and two FP7 EURATOM Projects: CINCH - a project formore » cooperation in nuclear chemistry - and ASGARD - a research project on advanced or novel nuclear fuels and their reprocessing issues for generation IV reactors. The integration of the university curricula in the market needs but also the anchoring in the research and future fuel cycles will be also discussed, with examples from the ASGARD project. (authors)« less

Modeling of selected ceramic processing parameters employed in the fabrication of 238PuO 2 fuel pellets

DOE PAGES

Brockman, R. A.; Kramer, D. P.; Barklay, C. D.; ...

2011-10-01

Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide ( 238PuO 2). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via ''classical'' ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processingmore » parameters with the goal of further enhancing the desired characteristics of the 238PuO 2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238PuO 2 itself has a significant thermal output. The results of the modeling efforts will be discussed.« less

Structural Health Monitoring of Nuclear Spent Fuel Storage Facilities

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

Yu, Lingyu

Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. To ensure that nuclear power remains clean energy, monitoring has been identified by DOE as a high priority cross-cutting need, necessary to determine and predict the degradation state of the systems, structures, and components (SSCs) important to safety (ITS). Therefore, nondestructive structural condition monitoring becomes a need to be installed on existing or to be integrated into future storage system to quantify the state of health or to guarantee the safe operation of nuclear power plants (NPPs) during their extendedmore » life span. In this project, the lead university and the collaborating national laboratory teamed to develop a nuclear structural health monitoring (n-SHM) system based on in-situ piezoelectric sensing technologies that can monitor structural degradation and aging for nuclear spent fuel DCSS and similar structures. We also aimed to identify and quantify possible influences of nuclear spent fuel environment (temperature and radiation) to the piezoelectric sensor system and come up with adequate solutions and guidelines therefore. We have therefore developed analytical model for piezoelectric based n-SHM methods, with considerations of temperature and irradiation influence on the model of sensing and algorithms in acoustic emission (AE), guided ultrasonic waves (GUW), and electromechanical impedance spectroscopy (EMIS). On the other side, experimentally the temperature and irradiation influence on the piezoelectric sensors and sensing capabilities were investigated. Both short-term and long-term irradiation investigation with our collaborating national laboratory were performed. Moreover, we developed multi-modal sensing, validated in laboratory setup, and conducted the testing on the We performed multi-modal sensing development, verification and validation tests on very complex

Exploration for fossil and nuclear fuels from orbital altitudes

NASA Technical Reports Server (NTRS)

Short, N. M.

1977-01-01

The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

Spent fuel cask handling at an operating nuclear power plant

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

Pal, A.C.

1988-01-01

The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices atmore » all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant.« less

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

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

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

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

Molten tin reprocessing of spent nuclear fuel elements

DOEpatents

Heckman, Richard A.

1983-01-01

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Economic Analysis of Complex Nuclear Fuel Cycles with NE-COST

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

Ganda, Francesco; Dixon, Brent; Hoffman, Edward

The purpose of this work is to present a new methodology, and associated computational tools, developed within the U.S. Department of Energy (U.S. DOE) Fuel Cycle Option Campaign to quantify the economic performance of complex nuclear fuel cycles. The levelized electricity cost at the busbar is generally chosen to quantify and compare the economic performance of different baseload generating technologies, including of nuclear: it is the cost of electricity which renders the risk-adjusted discounted net present value of the investment cash flow equal to zero. The work presented here is focused on the calculation of the levelized cost of electricitymore » of fuel cycles at mass balance equilibrium, which is termed LCAE (Levelized Cost of Electricity at Equilibrium). To alleviate the computational issues associated with the calculation of the LCAE for complex fuel cycles, a novel approach has been developed, which has been called the “island approach” because of its logical structure: a generic complex fuel cycle is subdivided into subsets of fuel cycle facilities, called islands, each containing one and only one type of reactor or blanket and an arbitrary number of fuel cycle facilities. A nuclear economic software tool, NE-COST, written in the commercial programming software MATLAB®, has been developed to calculate the LCAE of complex fuel cycles with the “island” computational approach. NE-COST has also been developed with the capability to handle uncertainty: the input parameters (both unit costs and fuel cycle characteristics) can have uncertainty distributions associated with them, and the output can be computed in terms of probability density functions of the LCAE. In this paper NE-COST will be used to quantify, as examples, the economic performance of (1) current Light Water Reactors (LWR) once-through systems; (2) continuous plutonium recycling in Fast Reactors (FR) with driver and blanket; (3) Recycling of plutonium bred in FR into LWR. For each

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Nuclear Fuel Reprocessing: U.S. Policy Development

DTIC Science & Technology

2006-11-29

to the chemical separation of fissionable uranium and plutonium from irradiated nuclear fuel. The World War II-era Manhattan Project developed...created the Atomic Energy Commission (AEC) and transferred production and control of fissionable materials from the Manhattan Project . As the exclusive

AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

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

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments weremore » transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.« less

Nuclear reactor fuel rod attachment system

DOEpatents

Not Available

1980-09-17

A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

Application of Compton-suppressed self-induced XRF to spent nuclear fuel measurement

NASA Astrophysics Data System (ADS)

Park, Se-Hwan; Jo, Kwang Ho; Lee, Seung Kyu; Seo, Hee; Lee, Chaehun; Won, Byung-Hee; Ahn, Seong-Kyu; Ku, Jeong-Hoe

2017-11-01

Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)'s spent nuclear fuel at the Korea atomic energy research institute (KAERI)'s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum's background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed selfinduced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

Support grid for fuel elements in a nuclear reactor

DOEpatents

Finch, Lester M.

1977-01-01

A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

Glass-ceramic material and method of making

DOEpatents

Meinhardt, Kerry D [Richland, WA; Vienna, John D [West Richland, WA; Armstrong, Timothy R [Pasco, WA; Pederson, Larry R [Kennewick, WA

2002-08-13

The present invention is a glass-ceramic material and method of making useful for joining at least two solid ceramic parts. The seal is a blend of M.sub.A O--M.sub.B O.sub.y --SiO.sub.2 that substantially matches a coefficient of thermal expansion of the solid electrolyte. According to the present invention, a series of glass ceramics in the M.sub.A O--M.sub.B O.sub.y --SiO.sub.2 system can be used to join or seal both tubular and planar ceramic solid oxide fuel cells, oxygen electrolyzers, and membrane reactors for the production of syngas, commodity chemicals and other products.

Nuclear core and fuel assemblies

DOEpatents

Downs, Robert E.

1981-01-01

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

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

2015-07-01

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

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

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

Hu, Jianwei; Gauld, Ian C.

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Seed and blanket fuel arrangement for dual-phase nuclear reactors

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

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

Methods and apparatuses for the development of microstructured nuclear fuels

DOEpatents

Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

2009-04-21

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

New generation nuclear fuel structures: Dense particles in selectively soluble matrix

NASA Astrophysics Data System (ADS)

Devlin, Dave; Jarvinen, Gordon; Patterson, Brian; Pattillo, Steve; Valdez, James; Liu, X.-Y.; Phillips, Jonathan

2009-11-01

We have developed a technology for dispersing sub-millimeter sized fuel particles within a bulk matrix that can be selectively dissolved. This may enable the generation of advanced nuclear fuels with easy separation of actinides and fission products. The large kinetic energy of the fission products results in most of them escaping from the sub-millimeter sized fuel particles and depositing in the matrix during burning of the fuel in the reactor. After the fuel is used and allowed to cool for a period of time, the matrix can be dissolved and the fission products removed for disposal while the fuel particles are collected by filtration for recycle. The success of such an approach would meet a major goal of the GNEP program to provide advanced recycle technology for nuclear energy production. The benefits of such an approach include (1) greatly reduced cost of the actinide/fission product separation process, (2) ease of recycle of the fuel particles, and (3) a radiation barrier to prevent theft or diversion of the recycled fuel particles during the time they are re-fabricated into new fuel. In this study we describe a method to make surrogate nuclear fuels of micrometer scale W (shell)/Mo (core) or HfO 2 particles embedded in an MgO matrix that allows easy separation of the fission products and their embedded particles. In brief, the method consists of physically mixing W-Mo or hafnia particles with an MgO precursor. Heating the mixture, in air or argon, without agitation, to a temperature is required for complete decomposition of the precursor. The resulting material was examined using chemical analysis, scanning electron microscopy, X-ray diffraction and micro X-ray computed tomography and found to consist of evenly dispersed particles in an MgO + matrix. We believe this methodology can be extended to actinides and other matrix materials.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

High temperature ceramic composition for hydrogen retention

DOEpatents

Webb, R.W.

1974-01-01

A ceramic coating for H retention in fuel elements is described. The coating has relatively low thermal neutron cross section, is not readily reduced by H at 1500 deg F, is adherent to the fuel element base metal, and is stable at reactor operating temperatures. (JRD)

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

NASA Astrophysics Data System (ADS)

Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

2017-05-01

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-03

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-155; 72-43 and NRC-2013-0218] Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory... the Big Rock Point (BRP) Independent Spent Fuel Storage Installation (ISFSI). ADDRESSES: Please refer...

Ceramic applications in the advanced Stirling automotive engine

NASA Technical Reports Server (NTRS)

Tomazic, W. A.; Cairelli, J. E.

1977-01-01

The ideal cycle, its application to a practical machine, and the specific advantages of high efficiency, low emissions, multi-fuel capability, and low noise of the stirling engine are discussed. Certain portions of the Stirling engine must operate continuously at high temperature. Ceramics offer the potential of cost reduction and efficiency improvement for advanced engine applications. Potential applications for ceramics in Stirling engines, and some of the special problems pertinent to using ceramics in the Stirling engine are described. The research and technology program in ceramics which is planned to support the development of advanced Stirling engines is outlined.

MicroRaman measurements for nuclear fuel reprocessing applications

DOE PAGES

Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

2016-12-01

Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO 3, and UO 2(NO 3) 2.« less

POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

DOEpatents

Dwyer, O.E.

1958-12-23

A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

Method for producing nuclear fuel

DOEpatents

Haas, Paul A.

1983-01-01

Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

77 FR 37937 - License Renewal Application for Prairie Island Nuclear Generating Plant Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

... Prairie Island Nuclear Generating Plant Independent Spent Fuel Storage Installation AGENCY: Nuclear... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... February 29, 2012 (ADAMS Accession number ML12065A073), by Prairie Island Nuclear Generating Plant (PINGP...

Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2017-01-01

To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

A method for monitoring nuclear absorption coefficients of aviation fuels

NASA Technical Reports Server (NTRS)

Sprinkle, Danny R.; Shen, Chih-Ping

1989-01-01

A technique for monitoring variability in the nuclear absorption characteristics of aviation fuels has been developed. It is based on a highly collimated low energy gamma radiation source and a sodium iodide counter. The source and the counter assembly are separated by a geometrically well-defined test fuel cell. A computer program for determining the mass attenuation coefficient of the test fuel sample, based on the data acquired for a preset counting period, has been developed and tested on several types of aviation fuel.

Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

NASA Astrophysics Data System (ADS)

Carlsen, Robert W.

Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

The role of accelerators in the nuclear fuel cycle

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

Takahashi, Hiroshi.

1990-01-01

The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the usemore » of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.« less

A combined gas cooled nuclear reactor and fuel cell cycle

NASA Astrophysics Data System (ADS)

Palmer, David J.

Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

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

Harrison, Thomas J.; Howard, Richard H.; Rader, Jordan D.

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examinationmore » facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.« less

Dissolution of used nuclear fuel using recycled nitric acid

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

Almond, Philip M.; Daniel, Jr., William E.; Rudisill, Tracy S.

An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.

Dissolution of used nuclear fuel using recycled nitric acid

DOE PAGES

Almond, Philip M.; Daniel, Jr., William E.; Rudisill, Tracy S.

2017-03-20

An evaluation was performed on the feasibility of using HB-Line anion exchange column waste streams from Alternate Feedstock 2 (AFS-2) processing for the dissolver solution for used nuclear fuel (UNF) processing. The targeted UNF for dissolution using recycled solution are fuels similar to the University of Missouri Research Reactor (MURR) fuel. Furthermore, the objectives of this experimental program were to validate the feasibility of using impure dissolver solutions with the MURR dissolution flowsheet to verify they would not significantly affect dissolution of the UNF in a detrimental manner.

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

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

Craft, Aaron E.; Papaioannou, Glen C.; Chichester, David L.

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutronmore » radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.« less

77 FR 59994 - Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-01

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-143; NRC-2009-0435] Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Notice of... Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

Ablation study of tungsten-based nuclear thermal rocket fuel

NASA Astrophysics Data System (ADS)

Smith, Tabitha Elizabeth Rose

The research described in this thesis has been performed in order to support the materials research and development efforts of NASA Marshall Space Flight Center (MSFC), of Tungsten-based Nuclear Thermal Rocket (NTR) fuel. The NTR was developed to a point of flight readiness nearly six decades ago and has been undergoing gradual modification and upgrading since then. Due to the simplicity in design of the NTR, and also in the modernization of the materials fabrication processes of nuclear fuel since the 1960's, the fuel of the NTR has been upgraded continuously. Tungsten-based fuel is of great interest to the NTR community, seeking to determine its advantages over the Carbide-based fuel of the previous NTR programs. The materials development and fabrication process contains failure testing, which is currently being conducted at MSFC in the form of heating the material externally and internally to replicate operation within the nuclear reactor of the NTR, such as with hot gas and RF coils. In order to expand on these efforts, experiments and computational studies of Tungsten and a Tungsten Zirconium Oxide sample provided by NASA have been conducted for this dissertation within a plasma arc-jet, meant to induce ablation on the material. Mathematical analysis was also conducted, for purposes of verifying experiments and making predictions. The computational method utilizes Anisimov's kinetic method of plasma ablation, including a thermal conduction parameter from the Chapman Enskog expansion of the Maxwell Boltzmann equations, and has been modified to include a tangential velocity component. Experimental data matches that of the computational data, in which plasma ablation at an angle shows nearly half the ablation of plasma ablation at no angle. Fuel failure analysis of two NASA samples post-testing was conducted, and suggestions have been made for future materials fabrication processes. These studies, including the computational kinetic model at an angle and the

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

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

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

Reactor-based management of used nuclear fuel: assessment of major options.

PubMed

Finck, Phillip J; Wigeland, Roald A; Hill, Robert N

2011-01-01

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms). Copyright © 2010 Health Physics Society

Mechanistic materials modeling for nuclear fuel performance

DOE PAGES

Tonks, Michael R.; Andersson, David; Phillpot, Simon R.; ...

2017-03-15

Fuel performance codes are critical tools for the design, certification, and safety analysis of nuclear reactors. However, their ability to predict fuel behavior under abnormal conditions is severely limited by their considerable reliance on empirical materials models correlated to burn-up (a measure of the number of fission events that have occurred, but not a unique measure of the history of the material). In this paper, we propose a different paradigm for fuel performance codes to employ mechanistic materials models that are based on the current state of the evolving microstructure rather than burn-up. In this approach, a series of statemore » variables are stored at material points and define the current state of the microstructure. The evolution of these state variables is defined by mechanistic models that are functions of fuel conditions and other state variables. The material properties of the fuel and cladding are determined from microstructure/property relationships that are functions of the state variables and the current fuel conditions. Multiscale modeling and simulation is being used in conjunction with experimental data to inform the development of these models. Finally, this mechanistic, microstructure-based approach has the potential to provide a more predictive fuel performance capability, but will require a team of researchers to complete the required development and to validate the approach.« less

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

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

IJ van Rooyen; WR Lloyd; TL Trowbridge

2013-09-01

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designsmore » being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between

Improving Prompt Temperature Feedback by Stimulating Doppler Broadening in Heterogeneous Composite Nuclear Fuel Forms

NASA Astrophysics Data System (ADS)

Morrison, Christopher

Nuclear fuels with similar aggregate material composition, but with different millimeter and micrometer spatial configurations of the component materials can have very different safety and performance characteristics. This research focuses on modeling and attempting to engineer heterogeneous combinations of nuclear fuels to improve negative prompt temperature feedback in response to reactivity insertion accidents. Improvements in negative prompt temperature feedback are proposed by developing a tailored thermal resistance in the nuclear fuel. In the event of a large reactivity insertion, the thermal resistance allows for a faster negative Doppler feedback by temporarily trapping heat in material zones with strong absorption resonances. A multi-physics simulation framework was created that could model large reactivity insertions. The framework was then used to model a comparison of a heterogeneous fuel with a tailored thermal resistance and a homogeneous fuel without the tailored thermal resistance. The results from the analysis confirmed the fundamental premise of prompt temperature feedback and provide insights into the neutron spectrum dynamics throughout the transient process. A trade study was conducted on infinite lattice fuels to help map a design space to study and improve prompt temperature feedback with many results. A multi-scale fuel pin analysis was also completed to study more realistic geometries. The results of this research could someday allow for novel nuclear fuels that would behave differently than current fuels. The idea of having a thermal barrier coating in the fuel is contrary to most current thinking. Inherent resistance to reactivity insertion accidents could enable certain reactor types once considered vulnerable to reactivity insertion accidents to be reevaluated in light of improved negative prompt temperature feedback.

High-Temperature, Thin-Film Ceramic Thermocouples Developed

NASA Technical Reports Server (NTRS)

Sayir, Ali; Blaha, Charles A.; Gonzalez, Jose M.

2005-01-01

To enable long-duration, more distant human and robotic missions for the Vision for Space Exploration, as well as safer, lighter, quieter, and more fuel efficient vehicles for aeronautics and space transportation, NASA is developing instrumentation and material technologies. The high-temperature capabilities of thin-film ceramic thermocouples are being explored at the NASA Glenn Research Center by the Sensors and Electronics Branch and the Ceramics Branch in partnership with Case Western Reserve University (CWRU). Glenn s Sensors and Electronics Branch is developing thin-film sensors for surface measurement of strain, temperature, heat flux, and surface flow in propulsion system research. Glenn s Ceramics Branch, in conjunction with CWRU, is developing structural and functional ceramic technology for aeropropulsion and space propulsion.

75 FR 51025 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-18

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle... meeting. SUMMARY: This notice announces an open meeting of the Reactor and Fuel Cycle Technology (RFCT... back end of the nuclear fuel cycle. The Commission will provide advice and make recommendations on...

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed

Apparatus for injection casting metallic nuclear energy fuel rods

DOEpatents

Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

1991-01-01

Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

Development of structural ceramic components for automobile applications

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

Kawamoto, H.

1995-12-01

Development efforts have been made in automobile technologies on heat engines to improve the power performance, the fuel economy, and so on. It is well recognized that ceramic applications are keys to succeed in such advanced heat engines, because of their good mechanical and thermal properties. This paper discusses present automobile applications of structural ceramic components and the expectations in automobile uses. The strength and reliability of mass-produced components for the engines are described with the manufacturing processes. The future R&D directions are recommended for structural ceramics.

Cold crucible induction melter test for crystalline ceramic waste form fabrication: A feasibility assessment

DOE PAGES

Amoroso, Jake W.; Marra, James; Dandeneau, Christopher S.; ...

2017-01-18

The first scaled proof-of-principle cold crucible induction melter (CCIM) test to process a multiphase ceramic waste form from a simulated combined (Cs/Sr, lanthanide and transition metal fission products) commercial used nuclear fuel waste stream was recently conducted in the United States. X-ray diffraction, 2-D X-ray absorption near edge structure (XANES), electron microscopy, inductively coupled plasma-atomic emission spectroscopy (and inductively coupled plasma-mass spectroscopy for Cs), and product consistency tests were used to characterize the fabricated CCIM material. Characterization analyses confirmed that a crystalline ceramic with a desirable phase assemblage was produced from a melt using a CCIM. We identified primary hollandite,more » pyrochlore/zirconolite, and perovskite phases in addition to minor phases rich in Fe, Al, or Cs. The material produced in the CCIM was chemically homogeneous and displayed a uniform phase assemblage with acceptable aqueous chemical durability.« less

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

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

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

Predicting Young’s Modulus of Glass/Ceramic Sealant for Solid Oxide Fuel Cell Considering the Combined Effects of Aging, Micro-Voids and Self-Healing

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

Liu, Wenning N.; Sun, Xin; Khaleel, Mohammad A.

We study the temperature dependent Young’s modulus for the glass/ceramic seal material used in Solid Oxide Fuel Cells (SOFCs). With longer heat treatment or aging time during operation, further devitrification may reduce the residual glass content in the seal material while boosting the ceramic crystalline content. In the meantime, micro-voids induced by the cooling process from the high operating temperature to room temperature can potentially degrade the mechanical properties of the glass/ceramic sealant. Upon reheating to the SOFC operating temperature, possible self-healing phenomenon may occur in the glass/ceramic sealant which can potentially restore some of its mechanical properties. A phenomenologicalmore » model is developed to model the temperature dependent Young’s modulus of glass/ceramic seal considering the combined effects of aging, micro-voids, and possible self-healing. An aging-time-dependent crystalline content model is first developed to describe the increase of the crystalline content due to the continuing devitrification under high operating temperature. A continuum damage mechanics (CDM) model is then adapted to model the effects of both cooling induced micro-voids and reheating induced self-healing. This model is applied to model the glass-ceramic G18, a candidate SOFC seal material previously developed at PNNL. Experimentally determined temperature dependent Young’s modulus is used to validate the model predictions« less

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

NASA Astrophysics Data System (ADS)

Åberg Lindell, M.; Andersson, P.; Grape, S.; Håkansson, A.; Thulin, M.

2018-07-01

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

DOE PAGES

Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

2016-01-01

This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systemsmore » on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.« less

Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

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

Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

2015-01-01

This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focused on issues of interest, e.g. the impact of the energy required to run the accelerator and associated systems onmore » the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are a critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also reviewed the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity versus a critical fast reactor with recycle of uranium and plutonium.« less

Accelerator-Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

NASA Astrophysics Data System (ADS)

Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

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

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Plutonium immobilization in glass and ceramics

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

Knecht, D.A.; Murphy, W.M.

1996-05-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposiummore » papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.« less

To Recycle or Not to Recycle? An Intergenerational Approach to Nuclear Fuel Cycles

PubMed Central

Kloosterman, Jan Leen

2007-01-01

This paper approaches the choice between the open and closed nuclear fuel cycles as a matter of intergenerational justice, by revealing the value conflicts in the production of nuclear energy. The closed fuel cycle improve sustainability in terms of the supply certainty of uranium and involves less long-term radiological risks and proliferation concerns. However, it compromises short-term public health and safety and security, due to the separation of plutonium. The trade-offs in nuclear energy are reducible to a chief trade-off between the present and the future. To what extent should we take care of our produced nuclear waste and to what extent should we accept additional risks to the present generation, in order to diminish the exposure of future generation to those risks? The advocates of the open fuel cycle should explain why they are willing to transfer all the risks for a very long period of time (200,000 years) to future generations. In addition, supporters of the closed fuel cycle should underpin their acceptance of additional risks to the present generation and make the actual reduction of risk to the future plausible. PMID:18075732

To recycle or not to recycle? An intergenerational approach to nuclear fuel cycles.

PubMed

Taebi, Behnam; Kloosterman, Jan Leen

2008-06-01

This paper approaches the choice between the open and closed nuclear fuel cycles as a matter of intergenerational justice, by revealing the value conflicts in the production of nuclear energy. The closed fuel cycle improve sustainability in terms of the supply certainty of uranium and involves less long-term radiological risks and proliferation concerns. However, it compromises short-term public health and safety and security, due to the separation of plutonium. The trade-offs in nuclear energy are reducible to a chief trade-off between the present and the future. To what extent should we take care of our produced nuclear waste and to what extent should we accept additional risks to the present generation, in order to diminish the exposure of future generation to those risks? The advocates of the open fuel cycle should explain why they are willing to transfer all the risks for a very long period of time (200,000 years) to future generations. In addition, supporters of the closed fuel cycle should underpin their acceptance of additional risks to the present generation and make the actual reduction of risk to the future plausible.

Demonstration of Subscale Cermet Fuel Specimen Fabrication Approach Using Spark Plasma Sintering and Diffusion Bonding

NASA Technical Reports Server (NTRS)

Barnes, Marvin W.; Tucker, Dennis S.; Benensky, Kelsa M.

2018-01-01

Nuclear thermal propulsion (NTP) has the potential to expand the limits of human space exploration by enabling crewed missions to Mars and beyond. The viability of NTP hinges on the development of a robust nuclear fuel material that can perform in the harsh operating environment (> or = 2500K, reactive hydrogen) of a nuclear thermal rocket (NTR) engine. Efforts are ongoing to develop fuel material and to assemble fuel elements that will be stable during the service life of an NTR. Ceramic-metal (cermet) fuels are being actively pursued by NASA Marshall Space Flight Center (MSFC) due to their demonstrated high-temperature stability and hydrogen compatibility. Building on past cermet fuel development research, experiments were conducted to investigate a modern fabrication approach for cermet fuel elements. The experiments used consolidated tungsten (W)-60vol%zirconia (ZrO2) compacts that were formed via spark plasma sintering (SPS). The consolidated compacts were stacked and diffusion bonded to assess the integrity of the bond lines and internal cooling channel cladding. The assessment included hot hydrogen testing of the manufactured surrogate fuel and pure W for 45 minutes at 2500 K in the compact fuel element environmental test (CFEET) system. Performance of bonded W-ZrO2 rods was compared to bonded pure W rods to access bond line integrity and composite stability. Bonded surrogate fuels retained structural integrity throughout testing and incurred minimal mass loss.

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

Nuclear reactor fuel rod attachment system

DOEpatents

Christiansen, David W.

1982-01-01

A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

Metals and Ceramics Division annual progress report, October 1, 1978-June 30, 1979

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

Peterson, S.

Research is reported concerning: (1) engineering materials including materials compatibility, mechanical properties, nondestructive testing, pressure vessel technology, and welding and brazing; (2) fuels and processes consisting of ceramic technology, fuel cycle technology, fuels evaluation, fuels fabrication and metals processing; and (3) materials science which includes, ceramic studies, physical metallurgy and properties, radiation effects and microstructural analysis, metastable and superconducting materials, structure and properties of surfaces, theoretical research, and x-ray research and applications. Highlights of the work of the metallographic group and the current status of the High-Temperature Materials Laboratory (HTML) and the Materials and Structures Technology Management Center (MSTMC) aremore » presented. (FS)« less

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

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

Adrian Miron; Joshua Valentine; John Christenson

2009-10-01

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFCmore » codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.« less

A statistical approach to nuclear fuel design and performance

NASA Astrophysics Data System (ADS)

Cunning, Travis Andrew

As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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

Hales, J. D.; Williamson, R. L.; Novascone, S. R.

2016-09-01

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact,more » and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.« less

Process Developed for Generating Ceramic Interconnects With Low Sintering Temperatures for Solid Oxide Fuel Cells

NASA Technical Reports Server (NTRS)

Zhong, Zhi-Min; Goldsby, Jon C.

2005-01-01

Solid oxide fuel cells (SOFCs) have been considered as premium future power generation devices because they have demonstrated high energy-conversion efficiency, high power density, and extremely low pollution, and have the flexibility of using hydrocarbon fuel. The Solid-State Energy Conversion Alliance (SECA) initiative, supported by the U.S. Department of Energy and private industries, is leading the development and commercialization of SOFCs for low-cost stationary and automotive markets. The targeted power density for the initiative is rather low, so that the SECA SOFC can be operated at a relatively low temperature (approx. 700 C) and inexpensive metallic interconnects can be utilized in the SOFC stack. As only NASA can, the agency is investigating SOFCs for aerospace applications. Considerable high power density is required for the applications. As a result, the NASA SOFC will be operated at a high temperature (approx. 900 C) and ceramic interconnects will be employed. Lanthanum chromite-based materials have emerged as a leading candidate for the ceramic interconnects. The interconnects are expected to co-sinter with zirconia electrolyte to mitigate the interface electric resistance and to simplify the processing procedure. Lanthanum chromites made by the traditional method are sintered at 1500 C or above. They react with zirconia electrolytes (which typically sinter between 1300 and 1400 C) at the sintering temperature of lanthanum chromites. It has been envisioned that lanthanum chromites with lower sintering temperatures can be co-fired with zirconia electrolyte. Nonstoichiometric lanthanum chromites can be sintered at lower temperatures, but they are unstable and react with zirconia electrolyte during co-sintering. NASA Glenn Research Center s Ceramics Branch investigated a glycine nitrate process to generate fine powder of the lanthanum-chromite-based materials. By simultaneously doping calcium on the lanthanum site, and cobalt and aluminum on the

Preparation, characterization and application of novel proton conducting ceramics

NASA Astrophysics Data System (ADS)

Wang, Siwei

Due to the immediate energy shortage and the requirement of environment protection nowadays, the efficient, effective and environmental friendly use of current energy sources is urgent. Energy conversion and storage is thus an important focus both for industry and academia. As one of the hydrogen energy related materials, proton conducting ceramics can be applied in solid oxide fuel cells and steam electrolysers, as well as high temperature hydrogen separation membranes and hydrogen sensors. For most of the practical applications, both high proton conductivity and chemical stability are desirable. However, the state-of-the-art proton conducting ceramics are facing great challenges in simultaneously fulfilling conductivity and stability requirements for practical applications. Consequently, understanding the properties for the proton conducting ceramics and developing novel materials that possess both high proton conductivity and enhanced chemical stability have both scientific and practical significances. The objective of this study is to develop novel proton conducting ceramics, either by evaluating the doping effects on the state-of-the-art simple perovskite structured barium cerates, or by investigating novel complex perovskite structured Ba3Ca1.18Nb1.82O 9-delta based proton conductors as potential proton conducting ceramics with improved proton conductivity and enhanced chemical stability. Different preparation methods were compared, and their influence on the structure, including the bulk and grain boundary environment has been investigated. In addition, the effects of microstructure on the electrical properties of the proton conducting ceramics have also been characterized. The solid oxide fuel cell application for the proton conducting ceramics performed as electrolyte membranes has been demonstrated.

Glass binder development for a glass-bonded sodalite ceramic waste form

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.; Canfield, Nathan L.; Zhu, Zihua; Zhang, Jiandong; Kruska, Karen; Schreiber, Daniel K.; Crum, Jarrod V.

2017-06-01

This paper discusses work to develop Na2O-B2O3-SiO2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. Here, five new glasses with ∼20 mass% Na2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. These improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.

Glass binder development for a glass-bonded sodalite ceramic waste form

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

Riley, Brian J.; Vienna, John D.; Frank, Steven M.

This paper discusses work to develop Na 2O-B 2O 3-SiO 2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na 2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion formore » the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.« less

Glass binder development for a glass-bonded sodalite ceramic waste form

DOE PAGES

Riley, Brian J.; Vienna, John D.; Frank, Steven M.; ...

2017-06-01

This paper discusses work to develop Na 2O-B 2O 3-SiO 2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na 2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion formore » the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.« less

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

DTIC Science & Technology

2010-03-05

However, the case of Iran raises perhaps the most critical question in this decade for strengthening the nuclear nonproliferation regime: How can...enrichment process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of...neutrons but can induce fission in all actinides , including all plutonium isotopes. Therefore, nuclear fuel for a fast reactor must have a higher

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

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

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Modeling and Simulation of a Nuclear Fuel Element Test Section

NASA Technical Reports Server (NTRS)

Moran, Robert P.; Emrich, William

2011-01-01

"The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

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

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light mostmore » of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.« less

DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

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

Sindelar, R.; Deible, R.

2011-04-27

The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report bymore » the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall

PROCESS OF MAKING A NEUTRONIC REACTOR FUEL ELEMENT COMPOSITION

DOEpatents

Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.

1959-01-13

A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.

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.

Electrospun Ceramic Nanofiber Mats Today: Synthesis, Properties, and Applications

PubMed Central

Esfahani, Hamid; Ramakrishna, Seeram

2017-01-01

Ceramic nanofibers (NFs) have recently been developed for advanced applications due to their unique properties. In this article, we review developments in electrospun ceramic NFs with regard to their fabrication process, properties, and applications. We find that surface activity of electrospun ceramic NFs is improved by post pyrolysis, hydrothermal, and carbothermal processes. Also, when combined with another surface modification methods, electrospun ceramic NFs result in the advancement of properties and widening of the application domains. With the decrease in diameter and length of a fiber, many properties of fibrous materials are modified; characteristics of such ceramic NFs are different from their wide and long (bulk) counterparts. In this article, electrospun ceramic NFs are reviewed with an emphasis on their applications as catalysts, membranes, sensors, biomaterials, fuel cells, batteries, supercapacitors, energy harvesting systems, electric and magnetic parts, conductive wires, and wearable electronic textiles. Furthermore, properties of ceramic nanofibers, which enable the above applications, and techniques to characterize them are briefly outlined. PMID:29077074

Electrospun Ceramic Nanofiber Mats Today: Synthesis, Properties, and Applications.