MRT fuel element inspection at Dounreay

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

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

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 support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.

PREIRRADIATION MEASUREMENTS OF PIQUA FUEL ELEMENTS NO. P-1111, P-1113, P- 1114, AND P-1120

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

Hubbell, H.J.

1962-11-01

Results of preirradiation measurements and tests performed during the processing and assembly of the individual fuel cylinders contained in Piqua Fuel Elements No. P-1111, P-1113, P-1114, and P-1120 are presented. A description of the techniques and equipment used in obtaining the data is also included. (auth)

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

2012-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

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

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

U-Mo Plate Blister Anneal Interim Report

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

Francine J. Rice; Daniel M. Wachs; Adam B. Robinson

2010-10-01

Blister thresholds in fuel elements have been a longstanding performance parameter for fuel elements of all types. This behavior has yet to be fully defined for the RERTR U-Mo fuel types. Blister anneal studies that began in 2007 have been expanded to include plates from more recent RERTR experiments. Preliminary data presented in this report encompasses the early generations of the U-Mo fuel systems and the most recent but still developing fuel system. Included is an overview of relevant dispersion fuel systems for the purposes of comparison.

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

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

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

Review of Rover fuel element protective coating development at Los Alamos

NASA Technical Reports Server (NTRS)

Wallace, Terry C.

1991-01-01

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

The Guardian: The Source for Antiterrorism Information. Volume 9, Number 1, April 2007

DTIC Science & Technology

2007-04-01

the fuel in these research reactors is generally not highly radioactive . Unlike the fuel rods in a nuclear power plant, these fuel elements would...NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT NUMBER 7. PERFORMING ORGANIZATION NAME(S) AND...practices and lessons learned. In addition, we will include Service and issue-specific breakout sessions that will focus on critical AT program elements

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

2013-01-01

To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen 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. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

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

Rest, J.; Hofman, G.L.

1997-06-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density.more » Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.« less

Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

2015-01-01

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at the time and resulted in NTREES being out of commission for a couple of months while a new stronger coil was procured. The new coil includes several additional pieces of support structure to prevent coil movement in the future. In addition, new insulating test article support components have been fabricated to prevent unexpected arcing to the test articles. Additional activities are also now underway to address ways in which the radial temperature profiles across test articles may be controlled such that they are more prototypical of what they would encounter in an operating nuclear engine. The causes of the temperature distribution problem are twofold. First, the fuel element test article is isolated in NTREES as opposed to being in the midst of many other mostly identical fuel elements in a nuclear engine. As a result, the fuel element heat flux boundary conditions in NTREES are far from adiabatic as would normally be the case in a reactor. Second, induction heating skews the power distribution such that power is preferentially deposited near the outside of the fuel element. Nuclear heating, conversely, deposits its power much more uniformly throughout the fuel element. Current studies are now looking at various schemes to adjust the amount of thermal radiation emitted from the fuel element surface so as to essentially vary the thermal boundary conditions on the test article. It is hoped that by properly adjusting the thermal boundary conditions on the fuel element test article, it may be possible to substantially correct for the inappropriate radial power distributions resulting from the induction heating so as to yield a more nearly correct temperature distribution throughout the fuel element.

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

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.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

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

Ruggles, A.E.

1990-10-12

The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results aremore » related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.« less

10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

Code of Federal Regulations, 2011 CFR

2011-01-01

... performance and safety during reactor operation. Also, in all cases precise control of processes, procedures... performance. (a) Items that are considered especially designed or prepared for the fabrication of fuel... pellets; (2) Automatic welding machines especially designed or prepared for welding end caps onto the fuel...

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie

2014-01-01

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

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

Sell, D. A.; Baily, C. E.; Malewitz, T. J.

2016-09-01

A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium aftermore » the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.« less

The Conceptual Design for a Fuel Assembly of a New Research Reactor

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

Ryu, J-S.; Cho, Y-G.; Yoon, D-B.

2004-10-06

A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibrationmore » characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.« less

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

Van Hagan, Thomas H.; Drees, Elizabeth A.

1993-01-01

Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

Code of Federal Regulations, 2010 CFR

2010-01-01

... performance and safety during reactor operation. Also, in all cases precise control of processes, procedures... elements include equipment that: (1) Normally comes in direct contact with, or directly processes or... pellets; (2) Automatic welding machines especially designed or prepared for welding end caps onto the fuel...

NTREES Testing and Operations Status

NASA Technical Reports Server (NTRS)

Emrich, Bill

2007-01-01

Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today's best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. Currently, the construction of such a simulator has been completed at the Marshall Space Flight Center, and will be used in the future to evaluate a wide variety of fuel element designs and the materials of which they are fabricated. This present work addresses the operational status of the Nuclear Thermal Rocket Element Environmental Simulator or NTREES and some of the design considerations which were considered prior to and during its construction.

The development of fuel performance models at the European institute for transuranium elements

NASA Astrophysics Data System (ADS)

Lassmann, K.; Ronchi, C.; Small, G. J.

1989-07-01

The design and operational performance of fuel rods for nuclear power stations has been the subject of detailed experimental research for over thirty years. In the last two decades the continuous demands for greater economy in conjunction with more stringent safety criteria have led to an increasing reliance on computer simulations. Conditions within a fuel rod must be calculated both for normal operation and for proposed reactor faults. It has thus been necessary to build up a reliable, theoretical understanding of the intricate physical, mechanical and chemical processes occurring under a wide range of conditions to obtain a quantitative insight into the behaviour of the fuel. A prime requirement, which has also proved to be the most taxing, is to predict the conditions under which failure of the cladding might occur, particularly in fuel nearing the end of its useful life. In this paper the general requirements of a fuel performance code are discussed briefly and an account is given of the basic concepts of code construction. An overview is then given of recent progress at the European Institute for Transuranium Elements in the development of a fuel rod performance code for general application and of more detailed mechanistic models for fission product behaviour.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

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

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« 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.

A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

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

Senor, David J.; Painter, Chad L.; Geelhood, Ken J.

2007-12-01

Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling,more » core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.« less

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

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

Temperature Profile in Fuel and Tie-Tubes for Nuclear Thermal Propulsion Systems

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

Vishal Patel

A finite element method to calculate temperature profiles in heterogeneous geometries of tie-tube moderated LEU nuclear thermal propulsion systems and HEU designs with tie-tubes is developed and implemented in MATLAB. This new method is compared to previous methods to demonstrate shortcomings in those methods. Typical methods to analyze peak fuel centerline temperature in hexagonal geometries rely on spatial homogenization to derive an analytical expression. These methods are not applicable to cores with tie-tube elements because conduction to tie-tubes cannot be accurately modeled with the homogenized models. The fuel centerline temperature directly impacts safety and performance so it must be predictedmore » carefully. The temperature profile in tie-tubes is also important when high temperatures are expected in the fuel because conduction to the tie-tubes may cause melting in tie-tubes, which may set maximum allowable performance. Estimations of maximum tie-tube temperature can be found from equivalent tube methods, however this method tends to be approximate and overly conservative. A finite element model of heat conduction on a unit cell can model spatial dependence and non-linear conductivity for fuel and tie-tube systems allowing for higher design fidelity of Nuclear Thermal Propulsion.« less

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David G; Chandler, David; Cook, David Howard

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

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

Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.

As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of themore » cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.« less

CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

DOEpatents

Weems, S.J.

1963-09-24

>A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

Multidimensional Fuel Performance Code: BISON

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

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

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.

MCNP-model for the OAEP Thai Research Reactor

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

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

Performance characteristics of LOX-H2, tangential-entry, swirl-coaxial, rocket injectors

NASA Technical Reports Server (NTRS)

Howell, Doug; Petersen, Eric; Clark, Jim

1993-01-01

Development of a high performing swirl-coaxial injector requires an understanding of fundamental performance characteristics. This paper addresses the findings of studies on cold flow atomic characterizations which provided information on the influence of fluid properties and element operating conditions on the produced droplet sprays. These findings are applied to actual rocket conditions. The performance characteristics of swirl-coaxial injection elements under multi-element hot-fire conditions were obtained by analysis of combustion performance data from three separate test series. The injection elements are described and test results are analyzed using multi-variable linear regression. A direct comparison of test results indicated that reduced fuel injection velocity improved injection element performance through improved propellant mixing.

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2014-01-01

Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit corrosion of certain potential NTR fuel forms. Additional diagnostic upgrades included in the present NTREES set up include the addition of a gamma ray spectrometer located near the vent filter to detect uranium fuel particles exiting the fuel element in the propellant exhaust stream to provide additional information any material loss occurring during testing. Other aspects of the upgrade included reworking NTREES to reduce the operational complexity of the system despite the increased complexity of the induction heating system. To this end, many of the controls were consolidated on fewer panels. As part of this upgrade activity, the Safety Assessment (SA) and the Standard Operating Procedures (SOPs) for NTREES were extensively rewritten. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can be accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements.

CFD-Based Design of a Filming Injector for N+3 Combustors

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2016-01-01

An effort was undertaken to perform CFD analysis of fluid flow in Lean-Direct Injection (LDI) combustors with axial swirl-venturi elements coupled with a new fuel-filming injector design for next-generation N+3 combustors. The National Combustion Code (NCC) was used to perform non-reacting and two-phase reacting flow computations on a N+3 injector configuration, in a single-element and a five-element injector array. All computations were performed with a consistent approach towards mesh-generation, spray-, ignition- and kinetics-modeling with the NCC. Computational predictions of the aerodynamics of the injector were used to arrive at an optimal injector design that met effective area, aerodynamics, and fuel-air mixing criteria. LDI-3 emissions (EINOx, EICO and UHC) were compared with the previous generation LDI-2 combustor experimental data at representative engine cycle conditions.

Modeling of Thermal Performance of Multiphase Nuclear Fuel Cell Under Variable Gravity Conditions

NASA Technical Reports Server (NTRS)

Ding, Z.; Anghaie, S.

1996-01-01

A unique numerical method has been developed to model the dynamic processes of bulk evaporation and condensation processes, associated with internal heat generation and natural convection under different gravity levels. The internal energy formulation, for the bulk liquid-vapor phase change problems in an encapsulated container, was employed. The equations, governing the conservation of mass, momentum and energy for both phases involved in phase change, were solved. The thermal performance of a multiphase uranium tetra-fluoride fuel element under zero gravity, micro-gravity and normal gravity conditions has been investigated. The modeling yielded results including the evolution of the bulk liquid-vapor phase change process, the evolution of the liquid-vapor interface, the formation and development of the liquid film covering the side wall surface, the temperature distribution and the convection flow field in the fuel element. The strong dependence of the thermal performance of such multiphase nuclear fuel cell on the gravity condition has been revealed. Under all three gravity conditions, 0-g, 10(exp -3)-g, and 1-g, the liquid film is formed and covers the entire side wall. The liquid film covering the side wall is more isothermalized at the wall surface, which can prevent the side wall from being over-heated. As the gravity increases, the liquid film is thinner, the temperature gradient is larger across the liquid film and smaller across the vapor phase. This investigation provides valuable information about the thermal performance of multi-phase nuclear fuel element for the potential space and ground applications.

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

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

1993-09-01

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

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

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

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

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

PubMed Central

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

2013-01-01

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

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.

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

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

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

FUEL ELEMENT SUPPORT

DOEpatents

Wyman, W.L.

1961-06-27

The described cylindrical fuel element has longitudinally spaced sets of short longitudinal ribs circumferentially spaced from one another. The ribs support the fuel element in a coolant tube so that there is an annular space for coolant flow between the fuel element and the interior of the coolant tube. If the fuel element grows as a result of reactor operation, the circumferential distribution of the ribs maintains the uniformity of the annular space between the coolant tube and the fuel element, and the collapsibility of the ribs prevents the fuel element from becoming jammed in the coolant tube.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David; Chandler, David; Cook, David

2014-10-30

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted usingmore » the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

Low temperature chemical processing of graphite-clad nuclear fuels

DOEpatents

Pierce, Robert A.

2017-10-17

A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Computational fluid dynamics analysis of SSME phase 2 and phase 2+ preburner injector element hydrogen flow paths

NASA Technical Reports Server (NTRS)

Ruf, Joseph H.

1992-01-01

Phase 2+ Space Shuttle Main Engine powerheads, E0209 and E0215 degraded their main combustion chamber (MCC) liners at a faster rate than is normal for phase 2 powerheads. One possible cause of the accelerated degradation was a reduction of coolant flow through the MCC. Hardware changes were made to the preburner fuel leg which may have reduced the resistance and, therefore, pulled some of the hydrogen from the MCC coolant leg. A computational fluid dynamics (CFD) analysis was performed to determine hydrogen flow path resistances of the phase 2+ fuel preburner injector elements relative to the phase 2 element. FDNS was implemented on axisymmetric grids with the hydrogen assumed to be incompressible. The analysis was performed in two steps: the first isolated the effect of the different inlet areas and the second modeled the entire injector element hydrogen flow path.

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

77 FR 5418 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

... aft fuel system 40 micron fuel filter element with a 10 micron fuel filter element. This proposed AD... fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. This proposed AD would require replacing each forward and aft fuel system 40 micron fuel filter element with a 10 micron fuel filter...

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

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

Wheelock, C.W.; Baumeister, E.B.

1961-09-01

A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

Low cost, lightweight fuel cell elements

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor)

2001-01-01

New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

Developmental status of thermionic materials.

NASA Technical Reports Server (NTRS)

Yang, L.; Chin, J.

1972-01-01

Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.

Performance of planar single cell lanthanum gallate based solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Maffei, N.; Kuriakose, A. K.

A novel synthesis of high purity, single phase strontium-magnesium doped lanthanum gallate through a nitrate route is described. The prepared powder is formed into planar monolithic elements by uniaxial pressing followed by isostatic pressing and sintering. XRD analysis of the sintered elements reveal no detectable secondary phases. The performance of the electrolyte in solid oxide fuel cells (SOFC) with three different anode/cathode combinations tested at 700°C with respect to the J- V and power density is reported. The data show that the characteristics of this SOFC are strongly dependent on the particular anode/cathode system chosen.

ORIGEN2 calculations supporting TRIGA irradiated fuel data package

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

Schmittroth, F.A.

ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.

RELAP5 Application to Accident Analysis of the NIST Research Reactor

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

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

NASA Astrophysics Data System (ADS)

Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

2014-01-01

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

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

Melissa Teague; Michael Tonks; Stephen Novascone

2014-01-01

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISONmore » fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.« less

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

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.

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

Requirements to the procedure and stages of innovative fuel development

NASA Astrophysics Data System (ADS)

Troyanov, V.; Zabudko, L.; Grachyov, A.; Zhdanova, O.

2016-04-01

According to the accepted current understanding under the nuclear fuel we will consider the assembled active zone unit (Fuel assembly) with its structural elements, fuel rods, pellet column, structural materials of fuel rods and fuel assemblies. The licensing process includes justification of safe application of the proposed modifications, including design-basis and experimental justification of the modified items under normal operating conditions and in violation of normal conditions, including accidents as well. Besides the justification of modified units itself, it is required to show the influence of modifications on the performance and safety of the other Reactor Unit’ and Nuclear Plant’ elements (e.g. burst can detection system, transportation and processing operations during fuel handling), as well as to justify the new standards of fuel storage etc. Finally, the modified fuel should comply with the applicable regulations, which often becomes a very difficult task, if only because those regulations, such as the NP-082-07, are not covered modification issues. Making amendments into regulations can be considered as the only solution, but the process is complicated and requires deep grounds for amendments. Some aspects of licensing new nuclear fuel are considered the example of mixed nitride uranium -plutonium fuel application for the BREST reactor unit.

The Manufacture of W-UO2 Fuel Elements for NTP Using the Hot Isostatic Pressing Consolidation Process

NASA Technical Reports Server (NTRS)

Broadway, Jeramie; Hickman, Robert; Mireles, Omar

2012-01-01

NTP is attractive for space exploration because: (1) Higher Isp than traditional chemical rockets (2)Shorter trip times (3) Reduced propellant mass (4) Increased payload. Lack of qualified fuel material is a key risk (cost, schedule, and performance). Development of stable fuel form is a critical path, long lead activity. Goals of this project are: Mature CERMET and Graphite based fuel materials and Develop and demonstrate critical technologies and capabilities.

Dissolution of Used Nuclear Fuel Using a TBP/N-Paraffin Solvent

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

Rudisill, T. S.; Shehee, T. C.; Jones, D. H.

2017-10-02

The dissolution of unirradiated used nuclear fuel (UNF) pellets pretreated for tritium removal was demonstrated using a tributly phosphate (TBP) solvent. Dissolution of pretreated fuel in TBP could potentially combine dissolution with two cycle of solvent extraction required for separating the actinides and lanthanides from other fission products. Dissolutions were performed using UNF surrogates prepared from both uranyl nitrate and uranium trioxide produced from the pretreatment process by adding selected actinide and stable fission product elements. In laboratory-scale experiments, the U dissolution efficiency ranged from 80-99+% for both the nitrate and oxide surrogate fuels. On average, 80% of the Pumore » and 50% of the Np and Am in the nitrate surrogate dissolved; however, little of the transuranic elements dissolved in the oxide form. The majority of the 3+ lanthanide elements dissolved. Only small amounts of Sr (0-1.6%) and Mo (0.1-1.7%) and essentially no Cs, Ru, Zr, or Pd dissolved.« less

CONCENTRIC TUBE FUEL ELEMENT SPRING ALIGNMENT SPACER DEVICE

DOEpatents

Weems, S.J.

1963-09-24

A rib construction for a nuclear-fuel element is described, in which one of three peripherally spaced ribs adjacent to each end of the fuel element is mounted on a radially yielding spring that embraces the fuel element. This spring enables the fuel element to have a good fit with a coolant tube and yet to be easily inserted in and withdrawn from the tube. (AEC)

Determination of the /sup 237/Np(n, 2n) reaction cross section in the core of the BN-350 reactor

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

Goncharov, R.K.; Zvonarev, A.V.; Ivanov, V.I.

1987-05-01

Analysis of irradiated fuel (regular fuel elements and special ampule samples) makes it possible to obtain data on the neutron-physical characteristics that do not lend themselves to measurement in experiments with a low neutron flux. The authors studied samples of spent fuel from regular fuel elements and samples of /sup 236/U and /sup 237/Np irradiated in special ampules. We give the results of measurements of the ratio (sigma/sub n,2n/ + sigma/sub ..gamma..,n/)sigma/sub c/ for /sup 237/Np in different zones of the BN-350 reactor and the results of calculations performed using the neptunium group constants calculated from the data of themore » files ENDFB IV and ENDFB V« less

A miniature fuel reformer system for portable power sources

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Belavič, Darko; Hrovat, Marko; Hočevar, Stanko; Pohar, Andrej; Petrovčič, Janko; Musizza, Bojan

2014-12-01

A miniature methanol reformer system has been designed and built to technology readiness level exceeding a laboratory prototype. It is intended to feed fuel cells with electric power up to 100 W and contains a complete setup of the technological elements: catalytic reforming and PROX reactors, a combustor, evaporators, actuation and sensing elements, and a control unit. The system is engineered not only for performance and quality of the reformate, but also for its lightweight and compact design, seamless integration of elements, low internal electric consumption, and safety. In the paper, the design of the system is presented by focussing on its miniaturisation, integration, and process control.

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

NASA Astrophysics Data System (ADS)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

NASA Technical Reports Server (NTRS)

Stewart, Mark E. M.

2015-01-01

This paper reviews test data for cermet fuel samples developed in the 1960's to better quantify Nuclear Thermal Propulsion (NTP) cermet engine performance, and to better understand contemporary fuel testing results. Over 200 cermet (W-UO2) samples were tested by thermally cycling to 2500 deg (2770 K) in hydrogen. The data indicates two issues at high temperatures: the vaporization rate of UO2 and the chemical stability of UO2. The data show that cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance, while other approaches yield smaller, incremental improvements. Data is very limited above 2770 K, and this complicates predictions of engine performance at high Isp. The paper considers how this material performance data translates into engine performance. In particular, the location of maximum temperature within the fuel element and the effect of heat deposition rate are examined.

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).

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)

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. ...

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

15. VIEW OF DUMMY FUEL ELEMENT ON FUEL ELEMENT HOLDER. SHOWS AIR FORCE MAN AT EDGE OF TANK. INEL PHOTO NUMBER 65-6176, TAKEN NOVEMBER 10, 1965. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Study of advanced fuel system concepts for commercial aircraft and engines

NASA Technical Reports Server (NTRS)

Versaw, E. F.; Brewer, G. D.; Byers, W. D.; Fogg, H. W.; Hanks, D. E.; Chirivella, J.

1983-01-01

The impact on a commercial transport aircraft of using fuels which have relaxed property limits relative to current commercial jet fuel was assessed. The methodology of the study is outlined, fuel properties are discussed, and the effect of the relaxation of fuel properties analyzed. Advanced fuel system component designs that permit the satisfactory use of fuel with the candidate relaxed properties in the subject aircraft are described. The two fuel properties considered in detail are freezing point and thermal stability. Three candidate fuel system concepts were selected and evaluated in terms of performance, cost, weight, safety, and maintainability. A fuel system that incorporates insulation and electrical heating elements on fuel tank lower surfaces was found to be most cost effective for the long term.

Energy Conversion Alternatives Study (ECAS), General Electric Phase 1. Volume 3: Energy conversion subsystems and components. Part 3: Gasification, process fuels, and balance of plant

NASA Technical Reports Server (NTRS)

Boothe, W. A.; Corman, J. C.; Johnson, G. G.; Cassel, T. A. V.

1976-01-01

Results are presented of an investigation of gasification and clean fuels from coal. Factors discussed include: coal and coal transportation costs; clean liquid and gas fuel process efficiencies and costs; and cost, performance, and environmental intrusion elements of the integrated low-Btu coal gasification system. Cost estimates for the balance-of-plant requirements associated with advanced energy conversion systems utilizing coal or coal-derived fuels are included.

Research on Fire-Resistant Diesel Fuel Flammability Mitigation Mechanisms

DTIC Science & Technology

1982-12-01

Naegeli U.S. Army Fuels and Lubricants Research Laboratory Southwest Research Institute San Antonio, Texas Under Contract to U.S. Army Mobility Equipment...David Naegeli DAAK70-80-C-0001 DAAK70-82-C-0001 9. PERFORMING ORGANIZATION NAME AND ADDRESSES 10. PROGRAM ELEMENT, PROJECT. TASK U.S. Army Fuels and...water-containing diesel fuel blends. 43 VI. LIST OF REFERENCES 1. Weatherford, W.D., Jr., Fodor, G.E., Naegeli , D.W., Wright, B.R., Owens, E.C., and

Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

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

Hanson, A. L.; Diamond, D.

2013-10-31

The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposedmore » LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.« less

Dismantlement of the TSF-SNAP Reactor Assembly

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

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated With Single/Dual-Spool Turbochargers

DTIC Science & Technology

2011-01-01

Comparative Performance Assessment of 5kW-Class Solid Oxide Fuel Cell Engines Integrated with Single/Dual-Spool Turbochargers So-Ryeok Oh, Jing Sun... Turbochargers 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER 5f. WORK UNIT... fundamental operating regime to the part load performance. Two different mechanical designs are assumed: dual shaft and single shaft as the compressor

NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Nuclear reactor

DOEpatents

Thomson, Wallace B.

2004-03-16

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

NASA Astrophysics Data System (ADS)

DeMange, P.; Marian, J.; Caro, M.; Caro, A.

2010-10-01

A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10 25 n m -2 ( E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Irradiation behavior of U 6Mn-Al dispersion fuel elements

NASA Astrophysics Data System (ADS)

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

2000-02-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

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

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

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

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

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE PAGES

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

2016-09-26

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

Fuel pumping system and method

DOEpatents

Shafer, Scott F [Morton, IL; Wang, Lifeng ,

2006-12-19

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Fuel Pumping System And Method

DOEpatents

Shafer, Scott F.; Wang, Lifeng

2005-12-13

A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

Basic elements of light water reactor fuel rod design. [FUELROD code

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

Weisman, J.; Eckart, R.

1981-06-01

Basic design techniques and equations are presented to allow students to understand and perform preliminary fuel design for normal reactor conditions. Each of the important design considerations is presented and discussed in detail. These include the interaction between fuel pellets and cladding and the changes in fuel and cladding that occur during the operating lifetime of the fuel. A simple, student-oriented, fuel rod design computer program, called FUELROD, is described. The FUELROD program models the in-pile pellet cladding interaction and allows a realistic exploration of the effect of various design parameters. By use of FUELROD, the student can gain anmore » appreciation of the fuel rod design process. 34 refs.« less

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

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

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

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

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

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

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

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

Fuel assembly for nuclear reactors

DOEpatents

Creagan, Robert J.; Frisch, Erling

1977-01-01

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

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.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

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.

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 4: High pressure fuel turbo-pump inlet housing analysis

NASA Technical Reports Server (NTRS)

Pool, Kirby V.

1989-01-01

The analysis performed on the Space Shuttle Main Engine (SSME) High Pressure Fuel Turbopump (HPFTP) inlet housings is summarized. Three DIAL finite element models were build to aid in assessing the structural life of the welds and fillets at the vanes. Complete results are given.

Current status of the development of high density LEU fuel for Russian research reactors

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

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Thermal-Hydraulic Transient Analysis of a Packed Particle Bed Reactor Fuel Element

DTIC Science & Technology

1990-06-01

long fuel elements, arranged to form a core , were analyzed for an up-power transient from 0 MWt to approximately 18 MWt. The simple model significantly...VARIATIONS IN FUEL ELEMENT GEOMETRY ............. 60 4.4 VARIATIONS IN THE MANNER OF TRANSIENT CONTROL ..... 62 4.5 CORE REPRESENTATION BY MULTIPLE FUEL ...the HTGR , however, the PBR packs small fuel particles between inner and outer retention elements, designated as frits. The PBR is appropriate for a

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.

78 FR 17591 - Airworthiness Directives; Sikorsky Aircraft Corporation Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-22

... aft fuel system 40 micron fuel filter element with a 10 micron nominal (40 micron absolute) fuel filter element. This AD was prompted by a National Transportation Safety Board (NTSB) review of in... helicopters with a fuel system 40 micron fuel filter element, part number (P/N) 52-0505-2 or 52-01064-1. That...

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Medical Isotope Production Analyses In KIPT Neutron Source Facility

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

Talamo, Alberto; Gohar, Yousry

Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is drivenmore » by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.« less

Production test IP-376-D, Supplement B Irradiation of MGCR-HDR-3 Test Element

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

Baars, R.E.

The objective of this supplement to PT-IP-376-D, Irradiation of MGCR-HDR-3 Test Element is to authorize 1000 hours of operation at a maximum test specimen surface temperature of 1700 F. The original production test authorized a test duration of four months at a maximum specimen surface temperature of 1500 F; supplement A authorized extension of the test duration to ten months. The desired increase in surface temperature is requested to demonstrate the general feasibility of operation of the fuel element at 1700 F, and to obtain specific information on the performance of Hastelloy-X cladding and fuel bodies. The increased temperature hasmore » been approved by the Atomic Energy Commission.« less

Effect of Air Swirler Configuration on Lean Direct Injector Flow Structure and Combustion Performance with a 7-Point Lean Direct Injector Array

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Tacina, Kathleen M.; Anderson, Robert C.

2017-01-01

This paper examines the fundamentals of fuel-air mixing in a lean direct injection concept. Results are presented to investigate the effects of air swirler angle, element spacing, and center element offset on recirculation zone formation, flame stability and gaseous emissions.

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-11-07

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Radial flow nuclear thermal rocket (RFNTR)

DOEpatents

Leyse, Carl F.

1995-01-01

A radial flow nuclear thermal rocket fuel assembly includes a substantially conical fuel element having an inlet side and an outlet side. An annular channel is disposed in the element for receiving a nuclear propellant, and a second, conical, channel is disposed in the element for discharging the propellant. The first channel is located radially outward from the second channel, and separated from the second channel by an annular fuel bed volume. This fuel bed volume can include a packed bed of loose fuel beads confined by a cold porous inlet frit and a hot porous exit frit. The loose fuel beads include ZrC coated ZrC-UC beads. In this manner, nuclear propellant enters the fuel assembly axially into the first channel at the inlet side of the element, flows axially across the fuel bed volume, and is discharged from the assembly by flowing radially outward from the second channel at the outlet side of the element.

Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Abdalla, Ayman

SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer Coefficient (HTC), bulk-fluid, sheath and fuel centerline temperature profiles were generated along the heated length of 5.772 m for a generic fuel channel. The fuel centerline and sheath temperature profiles have been determined at four Axial Heat Flux Profiles (AHFPs) using an average thermal power per channel of 8.5 MWth. The four examined AHFPs are the uniform, cosine, upstream-skewed and downstream-skewed profiles. Additionally, this study focuses on investigating a possibility of using low, enhanced and high thermal-conductivity fuels. The low thermal-conductivity fuels, which have been examined in this study, are uranium dioxide (UO 2), Mixed Oxide (MOX) and Thoria (ThO2) fuels. The examined enhanced thermal-conductivity fuels are uranium dioxide - silicon carbide (UO2 - SiC) and uranium dioxide - beryllium oxide (UO2 - BeO). Lastly, uranium carbide (UC), uranium dicarbide (UC2) and uranium nitride (UN) are the selected high thermal-conductivity fuels, which have been proposed for use in SCWRs. A comparison has been made between the low, enhanced and high thermal-conductivity fuels in order to identify the fuel centerline temperature behaviour when different nuclear fuels are used. Also, in the process of conducting the sensitivity analysis, the HTC was calculated using the Mokry et al. correlation, which is the most accurate supercritical water heat-transfer correlation so far. The sheath and the fuel centerline temperature profiles were determined for two cases. In Case 1, the HTC was calculated based on the Mokry et al. correlation, while in Case 2, the HTC values calculated for Case 1 were multiplied by a factor of 2. This factor was used in order to identify the amount of decrease in temperatures if the heat transfer is enhanced with appendages. Results of this analysis indicate that the use of the newly developed 54-element fuel bundle along with the proposed fuels is promising when compared with the Variant-20 (43-element) fuel bundle. Overall, the fuel centerline and sheath temperatures were below the industry and design limits when most of the proposed fuels were examined in the 54-element fuel bundle, however, the fuel centerline temperature limit was exceeded while MOX fuel was examined. Keywords: SCWRs, Fuel Centerline Temperature, Sheath Temperature, High Thermal Conductivity Fuels, Low Thermal Conductivity Fuels, HTC.

Gadolinia fuel performance in BWRs

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

Bailey, W.E.; Crowther, R.L.

1985-11-01

Gadolinia has the unique property of having a high neutron absorption cross section coupled with a burnup rate that can approximately match the uranium 235 depletion. These qualities and others make gadolinia an ideal burnable absorber, and it has been used in all General Electric-designed boiling water reactors. Fabrication corrosion properties, and performance of gadolinia-containing fuel elements are discussed. Development of a reliable and efficient set of local and global gadolinia-urania design methods has been an arduous process that has taken approx.15 years to accomplish.

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)

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

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

Wigner, E.P.; Young, G.J.; Weinberg, A.M.

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

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.

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.

SLUG HANDLING DEVICES

DOEpatents

Gentry, J.R.

1958-09-16

A device is described for handling fuel elements of a neutronic reactor. The device consists of two concentric telescoped contalners that may fit about the fuel element. A number of ratchet members, equally spaced about the entrance to the containers, are pivoted on the inner container and spring biased to the outer container so thnt they are forced to hear against and hold the fuel element, the weight of which tends to force the ratchets tighter against the fuel element. The ratchets are released from their hold by raising the inner container relative to the outer memeber. This device reduces the radiation hazard to the personnel handling the fuel elements.

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.

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.

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.

Durability test on irradiated rock-like oxide fuels

NASA Astrophysics Data System (ADS)

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

2003-06-01

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

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

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

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ...

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

35. DETAILS AND SECTIONS OF FUEL ELEMENT SUPPORT PLATFORM, FUEL ELEMENT HOLDER, TRIP MECHANISM COVER, AND OTHER DETAILS. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-3. INEL INDEX CODE NUMBER: 075 0701 60 851 151977. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

Possible consequences of operation with KIVN fuel elements in K Zircaloy process tubes

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

Carlson, P.A.

1963-08-06

From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy-2 process tube would be multiple fuel failure and failure of the process tube. The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The release of fission products would be expected to be no greater than that which has occurred followingmore » severe fuel failure incidents. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.« less

Exploring the Reduction of Fuel Consumption for Ship-to-Shore Connectors of the Marine Expeditionary Brigade

DTIC Science & Technology

2013-12-01

include the Command Element (CE), the Aviation Combat Element (ACE), the Ground Combat Element ( GCE ), and the Combat Service Support Element (CSSE) (from...35 Figure 9. The upper level functional architecture for EW12 shows the three primary functions of F1-Assemble the Seabase, F2...Function F2-Perform Assault is decomposed to obtain the lower level specific functions. From this decomposition, the sub-function of F2.2- Gain Entry

Crash Simulation of a Boeing 737 Fuselage Section Vertical Drop Test

NASA Technical Reports Server (NTRS)

Fasanella, Edwin L.; Jackson, Karen E.; Jones, Yvonne T.; Frings, Gary; Vu, Tong

2004-01-01

A 30-ft/s vertical drop test of a fuselage section of a Boeing 737 aircraft was conducted in October of 1999 at the FAA Technical Center in Atlantic City, NJ. This test was performed to evaluate the structural integrity of a conformable auxiliary fuel tank mounted beneath the floor and to determine its effect on the impact response of the airframe structure and the occupants. The test data were used to compare with a finite element simulation of the fuselage structure and to gain a better understanding of the impact physics through analytical/experimental correlation. To perform this simulation, a full-scale 3-dimensional finite element model of the fuselage section was developed using the explicit, nonlinear transient-dynamic finite element code, MSC.Dytran. The emphasis of the simulation was to predict the structural deformation and floor-level acceleration responses obtained from the drop test of the B737 fuselage section with the auxiliary fuel tank.

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.

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

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

Alternative aircraft fuels technology

NASA Technical Reports Server (NTRS)

Grobman, J.

1976-01-01

NASA is studying the characteristics of future aircraft fuels produced from either petroleum or nonpetroleum sources such as oil shale or coal. These future hydrocarbon based fuels may have chemical and physical properties that are different from present aviation turbine fuels. This research is aimed at determining what those characteristics may be, how present aircraft and engine components and materials would be affected by fuel specification changes, and what changes in both aircraft and engine design would be required to utilize these future fuels without sacrificing performance, reliability, or safety. This fuels technology program was organized to include both in-house and contract research on the synthesis and characterization of fuels, component evaluations of combustors, turbines, and fuel systems, and, eventually, full-scale engine demonstrations. A review of the various elements of the program and significant results obtained so far are presented.

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

DOEpatents

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

1973-12-11

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

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

Aghina, L.O.B.; Bellini, I.W.

The main works and modifications performed on the Argonaut reactor of the Instituto de Engenharia Nuclear after 4 years of operation are described. New positioning and holding system for the fuel elements and graphite wedges for the reactor brought appreciable stability to the operation. The removal of the plastic layer, printing and inspection of the corrosion of the elements plater are also described. (INIS)

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

DOE PAGES

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; ...

2016-01-01

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

454th Brookhaven Lecture

ScienceCinema

Charles Black

2017-12-09

Black discusses examples of integrating self-assembly into semiconductor microelectronics, where advances in the ability to define circuit elements at ever-higher resolution have largely fueled more than 40 years of consistent performance improvements

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

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.

Thermal finite-element analysis of space shuttle main engine turbine blade

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali; Tong, Michael T.; Kaufman, Albert

1987-01-01

Finite-element, transient heat transfer analyses were performed for the first-stage blades of the space shuttle main engine (SSME) high-pressure fuel turbopump. The analyses were based on test engine data provided by Rocketdyne. Heat transfer coefficients were predicted by performing a boundary-layer analysis at steady-state conditions with the STAN5 boundary-layer code. Two different peak-temperature overshoots were evaluated for the startup transient. Cutoff transient conditions were also analyzed. A reduced gas temperature profile based on actual thermocouple data was also considered. Transient heat transfer analyses were conducted with the MARC finite-element computer code.

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

Konyashov, Vadim V.; Krasnov, Alexander M.

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. Anmore » approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)« less

Affordable Development and Optimization of CERMET Fuels for NTP Ground Testing

NASA Technical Reports Server (NTRS)

Hickman, Robert R.; Broadway, Jeramie W.; Mireles, Omar R.

2014-01-01

CERMET fuel materials for Nuclear Thermal Propulsion (NTP) are currently being developed at NASA's Marshall Space Flight Center. The work is part of NASA's Advanced Space Exploration Systems Nuclear Cryogenic Propulsion Stage (NCPS) Project. The goal of the FY12-14 project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of an NTP system. A key enabling technology for an NCPS system is the fabrication of a stable high temperature nuclear fuel form. Although much of the technology was demonstrated during previous programs, there are currently no qualified fuel materials or processes. The work at MSFC is focused on developing critical materials and process technologies for manufacturing robust, full-scale CERMET fuels. Prototypical samples are being fabricated and tested in flowing hot hydrogen to understand processing and performance relationships. As part of this initial demonstration task, a final full scale element test will be performed to validate robust designs. The next phase of the project will focus on continued development and optimization of the fuel materials to enable future ground testing. The purpose of this paper is to provide a detailed overview of the CERMET fuel materials development plan. The overall CERMET fuel development path is shown in Figure 2. The activities begin prior to ATP for a ground reactor or engine system test and include materials and process optimization, hot hydrogen screening, material property testing, and irradiation testing. The goal of the development is to increase the maturity of the fuel form and reduce risk. One of the main accomplishmens of the current AES FY12-14 project was to develop dedicated laboratories at MSFC for the fabrication and testing of full length fuel elements. This capability will enable affordable, near term development and optimization of the CERMET fuels for future ground testing. Figure 2 provides a timeline of the development and optimization tasks for the AES FY15-17 follow on program.

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

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

NASA Astrophysics Data System (ADS)

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

2015-12-01

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

Conjugate Heat Transfer Analyses on the Manifold for Ramjet Fuel Injectors

NASA Technical Reports Server (NTRS)

Wang, Xiao-Yen J.

2006-01-01

Three-dimensional conjugate heat transfer analyses on the manifold located upstream of the ramjet fuel injector are performed using CFdesign, a finite-element computational fluid dynamics (CFD) software. The flow field of the hot fuel (JP-7) flowing through the manifold is simulated and the wall temperature of the manifold is computed. The three-dimensional numerical results of the fuel temperature are compared with those obtained using a one-dimensional analysis based on empirical equations, and they showed a good agreement. The numerical results revealed that it takes around 30 to 40 sec to reach the equilibrium where the fuel temperature has dropped about 3 F from the inlet to the exit of the manifold.

Developing and applying modern methods of leakage monitoring and state estimation of fuel at the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Povarov, V. P.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Pozychanyuk, I. V.; Gorobtsov, L. I.; Golubev, E. I.; Bykov, V. I.; Likhanskii, V. V.; Evdokimov, I. A.; Zborovskii, V. G.; Sorokin, A. A.; Kanyukova, V. D.; Aliev, T. N.

2014-02-01

The results of developing and implementing the modernized fuel leakage monitoring methods at the shut-down and running reactor of the Novovoronezh nuclear power plant (NPP) are presented. An automated computerized expert system integrated with an in-core monitoring system (ICMS) and installed at the Novovoronezh NPP unit no. 5 is described. If leaky fuel elements appear in the core, the system allows one to perform on-line assessment of the parameters of leaky fuel assemblies (FAs). The computer expert system units designed for optimizing the operating regimes and enhancing the fuel usage efficiency at the Novovoronezh NPP unit no. 5 are now being developed.

Water Chemistry Control System for Recovery of Damaged and Degraded Spent Fuel

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

Sindelar, R.; Fisher, D.; Thomas, J.

2011-02-18

The International Atomic Energy Agency (IAEA) and the government of Serbia have led the project cosponsored by the U.S, Russia, European Commission, and others to repackage and repatriate approximately 8000 spent fuel elements from the RA reactor fuel storage basins at the VIN?A Institute of Nuclear Sciences to Russia for reprocessing. The repackaging and transportation activities were implemented by a Russian consortium which includes the Sosny Company, Tekhsnabeksport (TENEX) and Mayak Production Association. High activity of the water of the fuel storage basin posed serious risk and challenges to the fuel removal from storage containers and repackaging for transportation. Themore » risk centered on personnel exposure, even above the basin water, due to the high water activity levels caused by Cs-137 leached from fuel elements with failed cladding. A team of engineers from the U.S. DOE-NNSA's Global Threat Reduction Initiative, the Vinca Institute, and the IAEA performed the design, development, and deployment of a compact underwater water chemistry control system (WCCS) to remove the Cs-137 from the basin water and enable personnel safety above the basin water for repackaging operations. Key elements of the WCCS system included filters, multiple columns containing an inorganic sorbent, submersible pumps and flow meters. All system components were designed to be remotely serviceable and replaceable. The system was assembled and successfully deployed at the Vinca basin to support the fuel removal and repackaging activities. Following the successful operations, the Cs-137 is now safely contained and consolidated on the zeolite sorbent used in the columns of the WCCS, and the fuel has been removed from the basins. This paper reviews the functional requirements, design, and deployment of the WCCS.« less

Photographic combustion characterization of LOX/Hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1980-01-01

One hundred twenty-seven tests were conducted over a chamber pressure range of 125-1500 psia, a fuel temperature range of -245 F to 158 F, and a fuel velocity range of 48-707 ft/sec to demonstrate the advantages and limitations of using high speed photography to identify potential combustion anomalies such as pops, fuel freezing, reactive stream separation and carbon formations. Combustion evaluation criteria were developed to guide selection of the fuels, injector elements, and operating conditions for testing. Separate criteria were developed for fuel and injector element selection and evaluation. The photographic test results indicated conclusively that injector element type and design directly influence carbon formation. Unlike spray fan, impingement elements reduce carbon formation because they induce a relatively rapid near zone fuel vaporization rate. Coherent jet impingement elements, on the other hand, exhibit increased carbon formation.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

NASA Astrophysics Data System (ADS)

Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, L.

2017-04-01

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

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.

Drying results of K-Basin fuel element 1990 (Run 1)

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

Marschman, S.C.; Abrefah, J.; Klinger, G.S.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

FUEL ELEMENT FOR NEUTRONIC REACTORS

DOEpatents

Evans, T.C.; Beasley, E.G.

1961-01-17

A fuel element for neutronic reactors, particularly the gas-cooled type of reactor, is described. The element comprises a fuel-bearing plate rolled to form a cylinder having a spiral passageway passing from its periphery to its center. In operation a coolant is admitted to the passageway at the periphery of the element, is passed through the spiral passageway, and emerges into a central channel defined by the inner turn of the rolled plate. The advantage of the element is that the fully heated coolant (i.e., coolant emerging into the central channel) is separated and thus insulated from the periphery of the element, which may be in contact with a low-temperature moderator, by the intermediate turns of the spiral fuel element.

Beam heated linear theta-pinch device for producing hot plasmas

DOEpatents

Bohachevsky, Ihor O.

1981-01-01

A device for producing hot plasmas comprising a single turn theta-pinch coil, a fast discharge capacitor bank connected to the coil, a fuel element disposed along the center axis of the coil, a predetermined gas disposed within the theta-pinch coil, and a high power photon, electron or ion beam generator concentrically aligned to the theta-pinch coil. Discharge of the capacitor bank generates a cylindrical plasma sheath within the theta-pinch coil which heats the outer layer of the fuel element to form a fuel element plasma layer. The beam deposits energy in either the cylindrical plasma sheath or the fuel element plasma layer to assist the implosion of the fuel element to produce a hot plasma.

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

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

Bruce G. Schnitzler; Stanley K. Borowski

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified asmore » the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.« less

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Schnitzler, Bruce G.; Borowski, Stanley K.

2012-01-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

Free Energy and Heat Capacity

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

Kurata, Masaki; Devanathan, Ramaswami

2015-10-13

Free energy and heat capacity of actinide elements and compounds are important properties for the evaluation of the safety and reliable performance of nuclear fuel. They are essential inputs for models that describe complex phenomena that govern the behaviour of actinide compounds during nuclear fuel fabrication and irradiation. This chapter introduces various experimental methods to measure free energy and heat capacity to serve as inputs for models and to validate computer simulations. This is followed by a discussion of computer simulation of these properties, and recent simulations of thermophysical properties of nuclear fuel are briefly reviewed.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

Babcock and Wilcox assessment of the Pratt and Whitney XNR2000

NASA Technical Reports Server (NTRS)

Westerman, Kurt O.; Scoles, Stephen W.; Jensen, R. R.; Rodes, J. R.; Ales, M. W.

1993-01-01

Babcock & Wilcox performed four subtasks related to the assessment of the Pratt & Whitney XNR2000 nuclear reactor as follows: (1) cermet fuel element fabricability assessment; (2) mechanical design review of the reactor system; (3) neutronic analysis review; and (4) safety assessment. The results of the mechanical and physics reviews have been integrated into the reactor design. The results of the fuel and safety assessments are presented.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

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

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Investigation of the Behavior of Fuel in the Intake Manifold and its Relation to S. I. Engines, 1980-1983

NASA Astrophysics Data System (ADS)

Servati, Hamid Beyragh

A liquid fuel film formation on the walls of an intake manifold adversely affects the engine performance and alters the overall air/fuel ratio from that scheduled by a fuel injector or carburetor and leads to adverse effects in vehicle driveability, exhaust emissions, and fuel economy. In this dissertation, the intake manifold is simulated by a horizontal circular duct. A model is provided to predict the rate of deposition and evaporation of the droplets in the intake manifold. The liquid fuel flow rate into the cylinders, mean film velocity and film thickness are determined as functions of engine parameters for both steady and transient operating conditions of the engine. A mathematical engine model is presented to simulate the dynamic interactions of the various engine components such as the air/fuel inlet element, intake manifold, combustion, dynamics and exhaust emissions. Inputs of the engine model are the intake manifold pressure and temperature, throttle angle, and air/fuel ratio. The observed parameters are the histories of fuel film thickness and velocity, fuel consumption, engine speed, engine speed hesitation time, and histories of CO, CO(,2), NO(,x), CH(,n), and O(,2). The effects of different air/fuel ratio control strategies on engine performance and observed parameters are also shown.

THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

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

Kittl, J.; Machado, R.E.; Mazza, J.A.

1958-06-10

The conditions required for the manufacture of the RA-1 Argonant type fuel elements are investigated. The fuel elements are in the form of a plate which is manufactured by the extrusion of a presintered mass of U/sub 3/O/sub 8/ (20% enriched) in an aluminum matrix. Steps in the investigation were obtention and specification of U/sub 3/O/sub 8/ and Al in powder form for testing, filling, and extrusion tests, finishing of the fuel elements, and computation of U/sub 3/O/sub 8/ content. (W.D.M.)

Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

NASA Technical Reports Server (NTRS)

Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

2013-01-01

In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

NASA Astrophysics Data System (ADS)

Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

2018-02-01

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

Co-Optimization of Fuels and Engines

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

Farrell, John

2016-03-24

The Co-Optimization of Fuels and Engines (Co-Optima) initiative is a new DOE initiative focused on accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) are designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, ten national laboratories, and numerous industry and academic partners to simultaneously tackle fuel and engine research and development (R&D) to maximize energymore » savings and on-road vehicle performance while dramatically reducing transportation-related petroleum consumption and greenhouse gas (GHG) emissions. This multi-year project will provide industry with the scientific underpinnings required to move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions. This presentation provides an overview of the project.« less

40 CFR 79.56 - Fuel and fuel additive grouping system.

Code of Federal Regulations, 2010 CFR

2010-07-01

... further testing under the provisions of Tier 3 or to support regulatory decisions affecting that fuel or... elements or classes of compounds other than those permitted in the base fuel for the respective fuel family... all of the following criteria: (1) Contain no elements other than carbon, hydrogen, oxygen, nitrogen...

Analysis and fabrication of tungsten CERMET materials for ultra-high temperature reactor applications via pulsed electric current sintering

NASA Astrophysics Data System (ADS)

Webb, Jonathan A.

The optimized development path for the fabrication of ultra-high temperature W-UO2 CERMET fuel elements were explored within this dissertation. A robust literature search was conducted, which concluded that a W-UO 2 fuel element must contain a fine tungsten microstructure and spherical UO2 kernels throughout the entire consolidation process. Combined Monte Carlo and Computational Fluid Dynamics (CFD) analysis were used to determine the effects of rhenium and gadolinia additions on the performance of W-UO 2 fuel elements at refractory temperatures and in dry and water submerged environments. The computational analysis also led to the design of quasi-optimized fuel elements that can meet thermal-hydraulic and neutronic requirements A rigorous set of experiments were conducted to determine if Pulsed Electric Current Sintering (PECS) can fabricate tungsten and W-Ce02 specimens to the required geometries, densities and microstructures required for high temperature fuel elements as well as determine the mechanisms involved within the PECS consolidation process. The CeO2 acts as a surrogate for UO 2 fuel kernels in these experiments. The experiments seemed to confirm that PECS consolidation takes place via diffusional mass transfer methods; however, the densification process is rapidly accelerated due to the effects of current densities within the consolidating specimen. Fortunately the grain growth proceeds at a traditional rate and the PECS process can yield near fully dense W and W-Ce02 specimens with a finer microstructure than other sintering techniques. PECS consolidation techniques were also shown to be capable of producing W-UO2 segments at near-prototypic geometries; however, great care must be taken to coat the fuel particles with tungsten prior to sintering. Also, great care must be taken to ensure that the particles remain spherical in geometry under the influence of a uniaxial stress as applied during PECS, which involves mixing different fuel kernel sizes in order to reduce the porosity in the initial green compact. Particle mixing techniques were also shown to be capable of producing consolidated CERMETs, but with a less than desirable microstructure. The work presented herin will help in the development of very high temperature reactors for terrestrial and space missions in the future.

Modelling of LOCA Tests with the BISON Fuel Performance Code

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

Williamson, Richard L; Pastore, Giovanni; Novascone, Stephen Rhead

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculationsmore » are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.« less

Compatibility Studies of Hydrogen Peroxide and a New Hypergolic Fuel Blend

NASA Technical Reports Server (NTRS)

Baldridge, Jennifer; Villegas, Yvonne

2002-01-01

Several preliminary materials compatibility studies have been conducted to determine the practicality of a new hypergolic fuel system. Hypergolic fuel ignites spontaneously as the oxidizer decomposes and releases energy in the presence of the fuel. The bipropellant system tested consists of high-test hydrogen peroxide (HTP) and a liquid fuel blend consisting of a hydrocarbon fuel, an ignition enhancer and a transition metal catalyst. In order for further testing of the new fuel blend to take place, some basic materials compatibility and HTP decomposition studies must be accomplished. The thermal decomposition rate of HTP was tested using gas evolution and isothermal microcalorimetry (IMC). Materials were analyzed for compatibility with hydrogen peroxide including a study of the affect welding has on stainless steel elemental composition and its relation to HTP decomposition. Compatibility studies of valve materials in the fuel blend were performed to determine the corrosion resistance of the materials.

Fuel elements of thermionic converters

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

Hunter, R.L.; Gontar, A.S.; Nelidov, M.V.

1997-01-01

Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving themore » following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.« less

The Benefits of Nuclear Thermal Propulsion (NTP) in an Evolvable Mars Campaign

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Mccurdy, David R.

2014-01-01

NTR: High thrust high specific impulse (2 x LOXLH2chemical) engine uses high power density fission reactor with enriched uranium fuel as thermal power source. Reactor heat is removed using H2propellant which is then exhausted to produce thrust. Conventional chemical engine LH2tanks, turbopumps, regenerative nozzles and radiation-cooled shirt extensions used --NTR is next evolutionary step in high performance liquid rocket engines During the Rover program, a common fuel element tie tube design was developed and used in the design of the 50 klbf Kiwi-B4E (1964), 75 klbf Phoebus-1B (1967), 250 klbf Phoebus-2A (June 1968), then back down to the 25 klbf Pewee engine (Nov-Dec 1968) NASA and DOE are using this same approach: design, build, ground then flight test a small engine using a common fuel element that is scalable to a larger 25 klbf thrust engine needed for human missions

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.

Low cost iodine intercalated graphene for fuel cells electrodes

NASA Astrophysics Data System (ADS)

Marinoiu, Adriana; Raceanu, Mircea; Carcadea, Elena; Varlam, Mihai; Stefanescu, Ioan

2017-12-01

On the theoretical predictions, we report the synthesis of iodine intercalated graphene for proton exchange membrane fuel cells (PEMFCs) applications. The structure and morphology of the samples were characterized by X-ray photoelectron spectroscopy (XPS) analysis, specific surface area by BET method, Raman investigations. The presence of elemental iodine in the form of triiodide and pentaiodide was validated, suggesting that iodine was trapped between graphene layers, leading to interactions with C atoms. The electrochemical performances of iodinated graphenes were tested and compared with a typical PEMFC configuration, containing different Pt/C loading (0.4 and 0.2 mg cm-2). If iodinated graphene is included as microporous layer, the electrochemical performances of the fuel cell are higher in terms of power density than the typical fuel cell. Iodine-doped graphenes have been successfully obtained by simple and cost effective synthetic strategy and demonstrated new insights for designing of a high performance metal-free ORR catalyst by a scalable technique.

Evaluation of Thin Plate Hydrodynamic Stability through a Combined Numerical Modeling and Experimental Effort

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

Tentner, A.; Bojanowski, C.; Feldman, E.

An experimental and computational effort was undertaken in order to evaluate the capability of the fluid-structure interaction (FSI) simulation tools to describe the deflection of a Missouri University Research Reactor (MURR) fuel element plate redesigned for conversion to lowenriched uranium (LEU) fuel due to hydrodynamic forces. Experiments involving both flat plates and curved plates were conducted in a water flow test loop located at the University of Missouri (MU), at conditions and geometries that can be related to the MURR LEU fuel element. A wider channel gap on one side of the test plate, and a narrower on the othermore » represent the differences that could be encountered in a MURR element due to allowed fabrication variability. The difference in the channel gaps leads to a pressure differential across the plate, leading to plate deflection. The induced plate deflection the pressure difference induces in the plate was measured at specified locations using a laser measurement technique. High fidelity 3-D simulations of the experiments were performed at MU using the computational fluid dynamics code STAR-CCM+ coupled with the structural mechanics code ABAQUS. Independent simulations of the experiments were performed at Argonne National Laboratory (ANL) using the STAR-CCM+ code and its built-in structural mechanics solver. The simulation results obtained at MU and ANL were compared with the corresponding measured plate deflections.« less

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

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

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.

Design and Fabrication of Oxygen/RP-2 Multi-Element Oxidizer-Rich Staged Combustion Thrust Chamber Injectors

NASA Technical Reports Server (NTRS)

Garcia, C. P.; Medina, C. R.; Protz, C. S.; Kenny, R. J.; Kelly, G. W.; Casiano, M. J.; Hulka, J. R.; Richardson, B. R.

2016-01-01

As part of the Combustion Stability Tool Development project funded by the Air Force Space and Missile Systems Center, the NASA Marshall Space Flight Center was contracted to assemble and hot-fire test a multi-element integrated test article demonstrating combustion characteristics of an oxygen/hydrocarbon propellant oxidizer-rich staged-combustion engine thrust chamber. Such a test article simulates flow through the main injectors of oxygen/kerosene oxidizer-rich staged combustion engines such as the Russian RD-180 or NK-33 engines, or future U.S.-built engine systems such as the Aerojet-Rocketdyne AR-1 engine or the Hydrocarbon Boost program demonstration engine. On the current project, several configurations of new main injectors were considered for the thrust chamber assembly of the integrated test article. All the injector elements were of the gas-centered swirl coaxial type, similar to those used on the Russian oxidizer-rich staged-combustion rocket engines. In such elements, oxidizer-rich combustion products from the preburner/turbine exhaust flow through a straight tube, and fuel exiting from the combustion chamber and nozzle regenerative cooling circuits is injected near the exit of the oxidizer tube through tangentially oriented orifices that impart a swirl motion such that the fuel flows along the wall of the oxidizer tube in a thin film. In some elements there is an orifice at the inlet to the oxidizer tube, and in some elements there is a sleeve or "shield" inside the oxidizer tube where the fuel enters. In the current project, several variations of element geometries were created, including element size (i.e., number of elements or pattern density), the distance from the exit of the sleeve to the injector face, the width of the gap between the oxidizer tube inner wall and the outer wall of the sleeve, and excluding the sleeve entirely. This paper discusses the design rationale for each of these element variations, including hydraulic, structural, thermal, combustion performance, and combustion stability considerations. This paper also discusses the fabrication and assembly of the injector components, including the injector body/interpropellant plate, the additive manufactured GRCop-84 faceplate, and the pieces that make up the injector elements including the oxidizer tube, an inlet to the oxidizer tube, and a facenut that includes the fuel tangential inlets and forms the initial recessed volume where oxidizer and fuel first interact. Hot-fire test results of these main injector designs in an integrated test article that includes an oxidizer-rich preburner are described in companion papers at this JANNAF meeting.

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

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

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Calculation of Distribution Dynamics of Inhomogeneous Temperature Field in Range of Fuel Elements by Using FreeFem++

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Shishkin, A. V.

2017-11-01

This article introduces the result of studying the heat exchange in the fuel element of the nuclear reactor fuel magazine. Fuel assemblies are completed as a bundle of cylindrical fuel elements located at the tops of a regular triangle. Uneven distribution of fuel rods in a nuclear reactor’s core forms the inhomogeneity of temperature fields. This article describes the developed method for heat exchange calculation with the account for impact of an inhomogeneous temperature field on the thermal-physical properties of materials and unsteady effects. The acquired calculation results are used for evaluating the tolerable temperature levels in protective case materials.

Removal of H{sub 2}S using molten carbonate at high temperature

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

Kawase, Makoto, E-mail: kawase@criepi.denken.or.jp; Otaka, Maromu

2013-12-15

Highlights: • The performance of molten carbonate for the removal of H{sub 2}S improves at higher temperatures. • The degree of H{sub 2}S removal is significantly affected by the CO{sub 2} concentration in syngas. • Addition of carbon elements, such as char and tar, decrease the negative effects of CO{sub 2}. • Continuous addition of carbon elements into molten carbonate enables continuous desulfurization. • Desulfurization using molten carbonate is suitable for gasification gas. - Abstract: Gasification is considered to be an effective process for energy conversion from various sources such as coal, biomass, and waste. Cleanup of the hot syngasmore » produced by such a process may improve the thermal efficiency of the overall gasification system. Therefore, the cleanup of hot syngas from biomass gasification using molten carbonate is investigated in bench-scale tests. Molten carbonate acts as an absorbent during desulfurization and dechlorination and as a thermal catalyst for tar cracking. In this study, the performance of molten carbonate for removing H{sub 2}S was evaluated. The temperature of the molten carbonate was set within the range from 800 to 1000 °C. It is found that the removal of H{sub 2}S is significantly affected by the concentration of CO{sub 2} in the syngas. When only a small percentage of CO{sub 2} is present, desulfurization using molten carbonate is inadequate. However, when carbon elements, such as char and tar, are continuously supplied, H{sub 2}S removal can be maintained at a high level. To confirm the performance of the molten carbonate gas-cleaning system, purified biogas was used as a fuel in power generation tests with a molten carbonate fuel cell (MCFC). The fuel cell is a high-performance sensor for detecting gaseous impurities. When purified gas from a gas-cleaning reactor was continuously supplied to the fuel cell, the cell voltage remained stable. Thus, the molten carbonate gas-cleaning reactor was found to afford good gas-cleaning performance.« less

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

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Fuel cell elements with improved water handling capacity

NASA Technical Reports Server (NTRS)

Kindler, Andrew (Inventor); Lee, Albany (Inventor)

2001-01-01

New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

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

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

NASA Astrophysics Data System (ADS)

Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

2017-04-01

Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Research requirements for development of regenerative engines for helicopters

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

Semple, R.D.

1976-12-01

The improved specific fuel consumption of the regenerative engine was compared to a simple-cycle turboshaft engine. The performance improvement and fuel saving are obtained at the expense of increased engine weight, development and production costs, and maintenance costs. Costs and schedules are estimated for the elements of the research and development program. Interaction of the regenerative engine with other technology goals for an advanced civil helicopter is examined, including its impact on engine noise, hover and cruise performance, helicopter empty weight, drive-system efficiency and weight, one-engine-inoperative hover capability, and maintenance and reliability.

Research requirements for development of regenerative engines for helicopters

NASA Technical Reports Server (NTRS)

Semple, R. D.

1976-01-01

The improved specific fuel consumption of the regenerative engine was compared to a simple-cycle turboshaft engine. The performance improvement and fuel saving are obtained at the expense of increased engine weight, development and production costs, and maintenance costs. Costs and schedules are estimated for the elements of the research and development program. Interaction of the regenerative engine with other technology goals for an advanced civil helicopter is examined, including its impact on engine noise, hover and cruise performance, helicopter empty weight, drive-system efficiency and weight, one-engine-inoperative hover capability, and maintenance and reliability.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

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

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

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

Gamble, Kyle Allan Lawrence; Novascone, Stephen Rhead; Gardner, Russell James

2016-06-01

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation andmore » power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-03-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Structural Evaluation of a Space Shuttle Main Engine (SSME) High Pressure Fuel Turbopump Turbine Blade

NASA Technical Reports Server (NTRS)

Abdul-Aziz, Ali

1996-01-01

Thermal and structural finite-element analyses were performed on the first high pressure fuel turbopump turbine blade of the space shuttle main engine (SSME). A two-dimensional (2-D) finite-element model of the blade and firtree disk attachment was analyzed using the general purpose MARC (finite-element) code. The loading history applied is a typical test stand engine cycle mission, which consists of a startup condition with two thermal spikes, a steady state and a shutdown transient. The blade material is a directionally solidified (DS) Mar-M 246 alloy, the blade rotor is forged with waspalloy material. Thermal responses under steady-state and transient conditions were calculated. The stresses and strains under the influence of mechanical and thermal loadings were also determined. The critical regions that exhibited high stresses and severe localized plastic deformation were the blade-rotor gaps.

Comparison of the properties of some synthetic crudes with petroleum crudes

NASA Technical Reports Server (NTRS)

Antoine, A. C.

1979-01-01

Physical properties and chemical compositions of six synthetic crudes were determined. The results were compared to those of typical petroleum crudes, with the interest being the feasibility of making jet fuels from oil shale and coal syncrudes. The specific gravity, viscosity, and pour point were measured, showing that these crudes would be described as heavier rather than lighter crudes. The boiling range distribution of the crudes was determined by distillation and by gas chromatography. In addition, gel permeation chromatograms were obtained, giving a unique molecular weight distribution profile for each crude. Analyses for carbon, hydrogen, nitrogen and sulfur concentrations were performed, as well as for hydrocarbon group type and trace element concentrations. It was found that the range in concentration of vanadium, an element whose presence in turbine fuels is of major concern, was lower than that of petroleum crudes. Sodium and potassium, other elements of concern, were present in comparatively high concentrations.

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.

Tag gas capsule with magnetic piercing device

DOEpatents

Nelson, Ira V.

1976-06-22

An apparatus for introducing a tag (i.e., identifying) gas into a tubular nuclear fuel element. A sealed capsule containing the tag gas is placed in the plenum in the fuel tube between the fuel and the end cap. A ferromagnetic punch having a penetrating point is slidably mounted in the plenum. By external electro-magnets, the punch may be caused to penetrate a thin rupturable end wall of the capsule and release the tag gas into the fuel element. Preferably the punch is slidably mounted within the capsule, which is in turn loaded as a sealed unit into the fuel element.

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

Barani, T.; Bruschi, E.; Pizzocri, D.

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

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

2006-06-01

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

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.

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

Direct carbon fuel cell and stack designs

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

Gorte, Raymond J.; Oh, Tae-Sik

Disclosed are novel configurations of Direct Carbon Fuel Cells (DCFCs), which optionally comprise a liquid anode. The liquid anode comprises a molten salt/metal, preferably Sb, and a fuel, which has significant elemental carbon content (coal, bio-mass, etc.). The supply of fuel is continuously replenished in the anode. In addition, a stack configuration is suggested where combining a large number of planar or tubular fuel elements.

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 6: Primary nozzle diffuser analysis

NASA Technical Reports Server (NTRS)

Foley, Michael J.

1989-01-01

The primary nozzle diffuser routes fuel from the main fuel valve on the Space Shuttle Main Engine (SSME) to the nozzle coolant inlet mainfold, main combustion chamber coolant inlet mainfold, chamber coolant valve, and the augmented spark igniters. The diffuser also includes the fuel system purge check valve connection. A static stress analysis was performed on the diffuser because no detailed analysis was done on this part in the past. Structural concerns were in the area of the welds because approximately 10 percent are in areas inaccessible by X-ray testing devices. Flow dynamics and thermodynamics were not included in the analysis load case. Constant internal pressure at maximum SSME power was used instead. A three-dimensional, finite element method was generated using ANSYS version 4.3A on the Lockheed VAX 11/785 computer to perform the stress computations. IDEAS Supertab on a Sun 3/60 computer was used to create the finite element model. Rocketdyne drawing number RS009156 was used for the model interpretation. The flight diffuser is denoted as -101. A description of the model, boundary conditions/load case, material properties, structural analysis/results, and a summary are included for documentation.

Nuclear breeder reactor fuel element with axial tandem stacking and getter

DOEpatents

Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

1981-01-01

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

Environmental Assessment for the Maintenance, Upgrade, and Construction of the Jet Fuel Distribution System, Edwards Air Force Base, California

DTIC Science & Technology

2009-01-01

5c. PROGRAM ELEMENT NUMBER 6. AUTHOR (S) 5d. PROJECT NUMBER 5e. TASK NUMBER EIAP Control No. 07-0485 5f. WORK UNIT NUMBER 7. PERFORMING...chemical hazards such as fuel vapors, heavy metal paints, and asbestos particulates; venomous snakes; and potential exposure to hantavirus or valley...OCCUPATIONAL HEALTH Environmental hazards Personnel could be exposed to heat stress, venomous snakes, hantavirus from infected rodents, and valley

A finite element analysis modeling tool for solid oxide fuel cell development: coupled electrochemistry, thermal and flow analysis in MARC®

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

Khaleel, Mohammad A.; Lin, Zijing; Singh, Prabhakar

2004-05-03

A 3D simulation tool for modeling solid oxide fuel cells is described. The tool combines the versatility and efficiency of a commercial finite element analysis code, MARC{reg_sign}, with an in-house developed robust and flexible electrochemical (EC) module. Based upon characteristic parameters obtained experimentally and assigned by the user, the EC module calculates the current density distribution, heat generation, and fuel and oxidant species concentration, taking the temperature profile provided by MARC{reg_sign} and operating conditions such as the fuel and oxidant flow rate and the total stack output voltage or current as the input. MARC{reg_sign} performs flow and thermal analyses basedmore » on the initial and boundary thermal and flow conditions and the heat generation calculated by the EC module. The main coupling between MARC{reg_sign} and EC is for MARC{reg_sign} to supply the temperature field to EC and for EC to give the heat generation profile to MARC{reg_sign}. The loosely coupled, iterative scheme is advantageous in terms of memory requirement, numerical stability and computational efficiency. The coupling is iterated to self-consistency for a steady-state solution. Sample results for steady states as well as the startup process for stacks with different flow designs are presented to illustrate the modeling capability and numerical performance characteristic of the simulation tool.« less

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

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

1961-01-01

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

CFD Evaluation of a 3rd Generation LDI Combustor

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2017-01-01

An effort was undertaken to perform CFD analysis of fluid flow in Lean-Direct Injection (LDI) combustors with axial swirl-venturi elements for next-generation LDI-3 combustor design. The National Combustion Code (NCC) was used to perform non-reacting and two-phase reacting flow computations for a nineteen-element injector array arranged in a three-module, 7-5-7 element configuration. All computations were performed with a consistent approach of mesh-optimization, spray-modeling, ignition and kinetics-modeling with the NCC. Computational predictions of the aerodynamics of the injector were used to arrive at an optimal injector design that meets effective area and fuel-air mixing criteria. LDI-3 emissions (EINOx, EICO and UHC) were compared with the previous generation LDI-2 combustor experimental data at representative engine cycle conditions.

Co-Optimization of Fuels and Engines

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

Farrell, John

2016-04-11

The Co-Optimization of Fuels and Engines (Co-Optima) initiative is a new DOE initiative focused on accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) are designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, ten national laboratories, and numerous industry and academic partners to simultaneously tackle fuel and engine research and development (R&D) to maximize energymore » savings and on-road vehicle performance while dramatically reducing transportation-related petroleum consumption and greenhouse gas (GHG) emissions. This multi-year project will provide industry with the scientific underpinnings required to move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions. This presentation provides an overview of the initiative and reviews recent progress focused on both advanced spark-ignition and compression-ignition approaches.« less

The Use Of Computational Human Performance Modeling As Task Analysis Tool

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

Jacuqes Hugo; David Gertman

2012-07-01

During a review of the Advanced Test Reactor safety basis at the Idaho National Laboratory, human factors engineers identified ergonomic and human reliability risks involving the inadvertent exposure of a fuel element to the air during manual fuel movement and inspection in the canal. There were clear indications that these risks increased the probability of human error and possible severe physical outcomes to the operator. In response to this concern, a detailed study was conducted to determine the probability of the inadvertent exposure of a fuel element. Due to practical and safety constraints, the task network analysis technique was employedmore » to study the work procedures at the canal. Discrete-event simulation software was used to model the entire procedure as well as the salient physical attributes of the task environment, such as distances walked, the effect of dropped tools, the effect of hazardous body postures, and physical exertion due to strenuous tool handling. The model also allowed analysis of the effect of cognitive processes such as visual perception demands, auditory information and verbal communication. The model made it possible to obtain reliable predictions of operator performance and workload estimates. It was also found that operator workload as well as the probability of human error in the fuel inspection and transfer task were influenced by the concurrent nature of certain phases of the task and the associated demand on cognitive and physical resources. More importantly, it was possible to determine with reasonable accuracy the stages as well as physical locations in the fuel handling task where operators would be most at risk of losing their balance and falling into the canal. The model also provided sufficient information for a human reliability analysis that indicated that the postulated fuel exposure accident was less than credible.« less

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 3B: High pressure fuel turbo-pump preburner pump bearing assembly analysis

NASA Technical Reports Server (NTRS)

Power, Gloria B.; Violett, Rebeca S.

1989-01-01

The analysis performed on the High Pressure Oxidizer Turbopump (HPOTP) preburner pump bearing assembly located on the Space Shuttle Main Engine (SSME) is summarized. An ANSYS finite element model for the inlet assembly was built and executed. Thermal and static analyses were performed.

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Priest, N D; Richardson, R B; Edwards, G W R

2013-02-01

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

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

Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier

The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimetermore » scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.« less

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.

High temperature fuel/emitter system for advanced thermionic fuel elements

NASA Astrophysics Data System (ADS)

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

1997-01-01

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

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

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

Rest, J.; Hofman, G.L.

1997-12-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data.

Reduced size fuel cell for portable applications

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor); Clara, Filiberto (Inventor); Frank, Harvey A. (Inventor)

2004-01-01

A flat pack type fuel cell includes a plurality of membrane electrode assemblies. Each membrane electrode assembly is formed of an anode, an electrolyte, and an cathode with appropriate catalysts thereon. The anode is directly into contact with fuel via a wicking element. The fuel reservoir may extend along the same axis as the membrane electrode assemblies, so that fuel can be applied to each of the anodes. Each of the fuel cell elements is interconnected together to provide the voltage outputs in series.

Coolant mass flow equalizer for nuclear fuel

DOEpatents

Betten, Paul R.

1978-01-01

The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

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

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

Fuel Cell Measurements with Cathode Catalysts of Sputtered Pt3 Y Thin Films.

PubMed

Lindahl, Niklas; Eriksson, Björn; Grönbeck, Henrik; Lindström, Rakel Wreland; Lindbergh, Göran; Lagergren, Carina; Wickman, Björn

2018-05-09

Fuel cells are foreseen to have an important role in sustainable energy systems, provided that catalysts with higher activity and stability are developed. In this study, highly active sputtered thin films of platinum alloyed with yttrium (Pt 3 Y) are deposited on commercial gas diffusion layers and their performance in a proton exchange membrane fuel cell is measured. After acid pretreatment, the alloy is found to have up to 2.5 times higher specific activity than pure platinum. The performance of Pt 3 Y is much higher than that of pure Pt, even if all of the alloying element was leached out from parts of the thin metal film on the porous support. This indicates that an even higher performance is expected if the structure of the Pt 3 Y catalyst or the support could be further improved. The results show that platinum alloyed with rare earth metals can be used as highly active cathode catalyst materials, and significantly reduce the amount of platinum needed, in real fuel cells. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

The combustion stability characteristics of a combustor consisting of a single shear element and a cylindrical chamber utilizing LOX and gaseous hydrogen as propellants are presented. The combustor geometry and the resulting longitudinal mode instability are axisymmetric. Hydrogen injection temperature and pyrotechnic pulsing were used to determine stability boundaries. Mixture ratio, fuel annulus gap, and LOX post configuration were varied. Performance and stability data were obtained for chamber pressures of 300 and 1000 psia.

Axisymmetric single shear element combustion instability experiment

NASA Technical Reports Server (NTRS)

Breisacher, Kevin J.

1993-01-01

The combustion stability characteristics of a combustor consisting of a single shear element and a cylindrical chamber utilizing LOX and gaseous hydrogen as propellants are presented. The combustor geometry and the resulting longitudinal mode instability are axisymmetric. Hydrogen injection temperature and pyrotechnic pulsing were used to determine stability boundaries. Mixture ratio, fuel annulus gap, and LOX post configuration were varied. Performance and stability data are presented for chamber pressures of 300 and 1000 psia.

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

PubMed

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

An Experimental Study of Ignition Effects and Flame Growth Over a Thin Solid Fuel in Low-Speed Concurrent Flow Using Drop-Tower Facilities

NASA Technical Reports Server (NTRS)

Pettegrew, Richard Dale

1996-01-01

An experimental study of ignition and flame growth over a thin solid fuel in oxidizer flow speeds from 0 to 10 cm/sec concurrent flow was performed. This study examined the differences between ignition using a resistively heated wire (woven in a sawtooth pattern over the leading edge of the fuel), and a straight resistively heated wire augmented by a chemical ignitor doped onto the leading edge of the fuel. Results showed that the chemical system yielded non-uniform ignition bursts, while the system using only the hotwire gave more uniform ignition. At speeds up to 2.5 cm/sec, the chemical system yielded non-uniform pyrolysis fronts, while the hotwire system gave more uniform pyrolysis fronts. At speeds of 5 cm/sec or greater, both systems gave uniform pyrolysis fronts. The chemically-ignited flames tended to become too dim to see faster than the hotwire-ignited flames, and the flame lengths were observed to be shorter (after the initial burst subsided) for the chemical system for all speeds. Flame and pyrolysis element velocities were measured. Temperature profiles for selected tests were measured using thermocouples at the fuel surface and in the gas phase. Comparisons between the flame element velocities and peak temperatures recorded in these tests with calculated spread rates and peak temperatures from a steady-state model are presented. Agreement was found to be within 20% for most flame elements for nominal velocities of 5 cm/sec and 7.5 cm/sec.

Improved nuclear fuel assembly grid spacer

DOEpatents

Marshall, John; Kaplan, Samuel

1977-01-01

An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

Prognosis and comparison of performances of composite CERCER and CERMET fuels dedicated to transmutation of TRU in an EFIT ADS

NASA Astrophysics Data System (ADS)

Sobolev, V.; Uyttenhove, W.; Thetford, R.; Maschek, W.

2011-07-01

The neutronic and thermomechanical performances of two composite fuel systems: CERCER with (Pu,Np,Am,Cm)O 2-x fuel particles in ceramic MgO matrix and CERMET with metallic Mo matrix, selected for transmutation of minor actinides in the European Facility for Industrial Transmutation (EFIT), were analysed aiming at their optimisation. The ALEPH burnup code system, based on MNCPX and ORIGEN codes and JEFF3.1 nuclear data library, and the modern version of the fuel rod performance code TRAFIC were used for this analysis. Because experimental data on the properties of the mixed minor-actinide oxides are scarce, and the in-reactor behaviour of the T91 steel chosen as cladding, as well as of the corrosion protective layer, is still not well-known, a set of "best estimates" provided the properties used in the code. The obtained results indicate that both fuel candidates, CERCER and CERMET, can satisfy the fuel design and safety criteria of EFIT. The residence time for both types of fuel elements can reach about 5 years with the reactivity swing within ±1000 pcm, and about 22% of the loaded MA is transmuted during this period. However, the fuel centreline temperature in the hottest CERCER fuel rod is close to the temperature above which MgO matrix becomes chemically instable. Moreover, a weak PCMI can appear in about 3 years of operation. The CERMET fuel can provide larger safety margins: the fuel temperature is more than 1000 K below the permitted level of 2380 K and the pellet-cladding gap remains open until the end of operation.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

JACKETED FUEL ELEMENT

DOEpatents

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

1959-02-01

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

Comprehensive Fuel Spray Modeling and Impacts on Chamber Acoustics in Combustion Dynamics Simulations

DTIC Science & Technology

2013-05-01

multiple swirler configurations and fuel injector locations at atmospheric pressure con- ditions. Both single-element and multiple-element LDI...the swirl number, Reynolds’ number and injector location in the LDI element. Besides the multi-phase flow characteristics, several experimen- tal...region downstream of the fuel injector on account of a sta- ble and compact precessing vortex core. Recent ex- periments conducted by the Purdue group have

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.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Experimental design of an interlaboratory study for trace metal analysis of liquid fluids. [for aerospace vehicles

NASA Technical Reports Server (NTRS)

Greenbauer-Seng, L. A.

1983-01-01

The accurate determination of trace metals and fuels is an important requirement in much of the research into and development of alternative fuels for aerospace applications. Recognizing the detrimental effects of certain metals on fuel performance and fuel systems at the part per million and in some cases part per billion levels requires improved accuracy in determining these low concentration elements. Accurate analyses are also required to ensure interchangeability of analysis results between vendor, researcher, and end use for purposes of quality control. Previous interlaboratory studies have demonstrated the inability of different laboratories to agree on the results of metal analysis, particularly at low concentration levels, yet typically good precisions are reported within a laboratory. An interlaboratory study was designed to gain statistical information about the sources of variation in the reported concentrations. Five participant laboratories were used on a fee basis and were not informed of the purpose of the analyses. The effects of laboratory, analytical technique, concentration level, and ashing additive were studied in four fuel types for 20 elements of interest. The prescribed sample preparation schemes (variations of dry ashing) were used by all of the laboratories. The analytical data were statistically evaluated using a computer program for the analysis of variance technique.

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

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

Multidimensional Multiphysics Simulation of TRISO Particle Fuel

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

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

2013-11-01

Multidimensional multiphysics analysis of TRISO-coated particle fuel using the BISON finite-element based nuclear fuels code is described. The governing equations and material models applicable to particle fuel and implemented in BISON are outlined. Code verification based on a recent IAEA benchmarking exercise is described, and excellant comparisons are reported. Multiple TRISO-coated particles of increasing geometric complexity are considered. It is shown that the code's ability to perform large-scale parallel computations permits application to complex 3D phenomena while very efficient solutions for either 1D spherically symmetric or 2D axisymmetric geometries are straightforward. Additionally, the flexibility to easily include new physical andmore » material models and uncomplicated ability to couple to lower length scale simulations makes BISON a powerful tool for simulation of coated-particle fuel. Future code development activities and potential applications are identified.« less

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

Potter, David Charles; Taylor, Craig Michael; Coons, James Elmer

The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM fuel samples because the spectra was recorded from the source in very close proximity to the detector and surrounded by the guard annulus, so the detection probability is very high. Though this configuration is optimal for a Compton suppression system for the measurement of low count rate samples, measurement of high count rate samples in the enclosed arrangement leads to sum peaks in both the suppressed and unsuppressed spectra and losses to photopeak counts in the suppressed spectra. No additional photopeaks were detected using Compton suppression with this geometry. A detector model was constructed that can accurately simulate a Compton suppressed spectral measurement of radiation from spent nuclear fuel using HPGe or LaBr3 detectors. This is the first detector model capable of such an accomplishment. The model uses the Geant4 toolkit coupled with the RadSrc application and it accepts spent fuel composition data in list form. The model has been validated using dissolved ATM fuel samples in the standard, enclosed geometry of the PSU HPGe-CSS. The model showed generally good agreement with both the unsuppressed and suppressed measured fuel sample spectra, however the simulation is more appropriate for the generation of gamma-ray spectra in the beam source configuration. Photopeak losses due to cascade decay emissions in the Compton suppressed spectra were not appropriately managed by the simulation. Compton suppression would be a beneficial addition to NDA process monitoring systems if oriented such that the gamma-ray photons are collimated to impinge the primary detector face as a beam. The analysis has shown that peak losses through accidental coincidences are minimal and the reduction in the Compton continuum allows additional peaks to be resolved. (Abstract shortened by UMI.).

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

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

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

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

2015-09-01

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

Test-Analysis Correlation of a Crash Simulation of a Vertical Drop Test of a Commuter-Category Aircraft

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.

2004-01-01

A finite element model of an ATR42-300 commuter-class aircraft was developed and a crash simulation was executed. Analytical predictions were correlated with data obtained from a 30-feet per second (9.14-meters per second) vertical drop test of the aircraft. The purpose of the test was to evaluate the structural response of the aircraft when subjected to a severe, but survivable, impact. The aircraft was configured with seats, dummies, luggage, and other ballast. The wings were filled with 8,700 lb. (3,946 kilograms) of water to represent the fuel. The finite element model, which consisted of 57,643 nodes and 62,979 elements, was developed from direct measurements of the airframe geometry. The seats, dummies, luggage, simulated engines and fuel, and other ballast were represented using concentrated masses. The model was executed in LS-DYNA, a commercial finite element code for performing explicit transient dynamic simulations. Analytical predictions of structural deformation and selected time-history responses were correlated with experimental data from the drop test to validate the simulation.

Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

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

Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean

2010-04-01

The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction processmore » was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.« less

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

DOE PAGES

Barani, T.; Bruschi, E.; Pizzocri, D.; ...

2017-01-03

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which ismore » applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.« less

Fuel shipment experience, fuel movements from the BMI-1 transport cask

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

Bauer, Thomas L.; Krause, Michael G

1986-07-01

The University of Texas at Austin received two shipments of irradiated fuel elements from Northrup Aircraft Corporation on April 11 and 16, 1985. A total of 59 elements consisting of standard and instrumented TRIGA fuel were unloaded from the BMI-1 shipping cask. At the time of shipment, the Northrup core burnup was approximately 50 megawatt days with fuel element radiation levels, after a cooling time of three months, of approximately 1.75 rem/hr at 3 feet. In order to facilitate future planning of fuel shipment at the UT facility and other facilities, a summary of the recent transfer process including severalmore » factors which contributed to its success are presented. Numerous color slides were made of the process for future reference by UT and others involved in fuel transfer and handling of the BMI-1 cask.« less

The Finite Element Modelling and Dynamic Characteristics Analysis about One Kind of Armoured Vehicles’ Fuel Tanks

NASA Astrophysics Data System (ADS)

Gao, Yang; Ge, Zhishang; Zhai, Weihao; Tan, Shiwang; Zhang, Feng

2018-01-01

The static and dynamic characteristics of fuel tank are studied for the armoured vehicle in this paper. The CATIA software is applied to build the CAD model of the armoured vehicles’ fuel tank, and the finite element model is established in ANSYS Workbench. The finite element method is carried out to analyze the static and dynamic mechanical properties of the fuel tank, and the first six orders of mode shapes and their frequencies are also computed and given in the paper, then the stress distribution diagram and the high stress areas are obtained. The results of the research provide some references to the fuel tanks’ design improvement, and give some guidance for the installation of the fuel tanks on armoured vehicles, and help to improve the properties and the service life of this kind of armoured vehicles’ fuel tanks.

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

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work inmore » the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.« less

Co-Optimization of Fuels & Engines for Tomorrow's Energy-Efficient Vehicles

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

2016-03-01

A new U.S. Department of Energy (DOE) initiative is accelerating the introduction of affordable, scalable, and sustainable biofuels and high-efficiency, low-emission vehicle engines. The simultaneous fuels and vehicles research and development (R&D) is designed to deliver maximum energy savings, emissions reduction, and on-road vehicle performance. The initiative's integrated approach combines the previously independent areas of biofuels and combustion R&D, bringing together two DOE Office of Energy Efficiency & Renewable Energy research offices, nine national laboratories, and numerous industry and academic partners to more rapidly identify commercially viable solutions. This multi-year project will provide industry with the scientific underpinnings required tomore » move new biofuels and advanced engine systems to market faster while identifying and addressing barriers to their commercialization. This project's ambitious, first-of-its-kind approach simultaneously tackles fuel and engine innovation to co-optimize performance of both elements and provide dramatic and rapid cuts in fuel use and emissions.« less

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

NASA Astrophysics Data System (ADS)

Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

2004-07-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

PROTECTIVELY COVERED ARTICLE AND METHOD OF MANUFACTURE

DOEpatents

Plott, R.F.

1958-10-28

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

URANIUM OXIDE-CONTAINING FUEL ELEMENT COMPOSITION AND METHOD OF MAKING SAME

DOEpatents

Handwerk, J.H.; Noland, R.A.; Walker, D.E.

1957-09-10

In the past, bodies formed of a mixture of uranium dioxide and aluminum powder have been used in fuel elements; however, these mixtures were found not to be suitable when exposed to temperatures of about 600 deg C, because at such high temperatures the fuel elements were distorted. If uranosic oxide, U/sub 3/O/sub 8/, is substituted for UO/sub 2/, the mechanical properties are not impaired when these materials are used at about 600 deg C and no distortion takes place. The uranosic oxide and aluminum, both in powder form, are first mixed, and after a homogeneous mixture has been obtained, are shaped into fuel elements by extrusion at elevated temperature. Magnesium powder may be used in place of the aluminum.

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

JACKETED FISSIONABLE MEMBER

DOEpatents

Boller, E.R.; Robinson, J.W.

1960-09-13

A fuel element design for a nuclear reactor is presented. The fuel element comprises a cylindrical fuel body having a portion of smaller diameter at each end thereof with an annular flange at the extreme ends of these portions of smaller diameter. An end cap fits over the ends of the fuel body and has an internal annular groove adapted to receive the flange. The fuel body and end caps are disposed in a cup-shaped jacket, a closure disc completing the enclosure of the fuel body, and tht caps are bonded over their entire periphery to the jacket.

Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

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

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (aboutmore » 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)« less

Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

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

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu andmore » Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)« less

Turning Perspective in Photoelectrocatalytic Cells for Solar Fuels.

PubMed

Perathoner, Siglinda; Centi, Gabriele; Su, Dangsheng

2016-02-19

The development of new devices for the use and storage of solar energy is a key step to enable a new sustainable energy scenario. The route for direct solar-to-chemical energy transformation, especially to produce liquid fuels, represents a necessary element to realize transition from the actual energy infrastructure. Photoelectrocatalytic (PECa) devices for the production of solar fuels are a key element to enable this sustainable scenario. The development of PECa devices and related materials is of increasing scientific and applied interest. This concept paper introduces the need to turn the viewpoint of research in terms of PECa cell design and related materials with respect to mainstream activities in the field of artificial photosynthesis and leaves. As an example of a new possible direction, the concept of electrolyte-less cell design for PECa cells to produce solar fuels by reduction of CO2 is presented. The fundamental and applied development of new materials and electrodes for these cells should proceed fully integrated with PECa cell design and systematic analysis. A new possible approach to develop semiconductors with improved performances by using visible light is also shortly presented. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Petrucco, Louis Jacob (Inventor); Daum, Wolfgang (Inventor)

2005-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

2003-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Fiber optic sensors for gas turbine control

NASA Technical Reports Server (NTRS)

Shu, Emily Yixie (Inventor); Brown, Dale Marius (Inventor); Petrucco, Louis Jacob (Inventor); Lovett, Jeffery Allan (Inventor); Daum, Wolfgang (Inventor); Dunki-Jacobs, Robert John (Inventor)

1999-01-01

An apparatus for detecting flashback occurrences in a premixed combustor system having at least one fuel nozzle includes at least one photodetector and at least one fiber optic element coupled between the at least one photodetector and a test region of the combustor system wherein a respective flame of the fuel nozzle is not present under normal operating conditions. A signal processor monitors a signal of the photodetector. The fiber optic element can include at least one optical fiber positioned within a protective tube. The fiber optic element can include two fiber optic elements coupled to the test region. The optical fiber and the protective tube can have lengths sufficient to situate the photodetector outside of an engine compartment. A plurality of fuel nozzles and a plurality of fiber optic elements can be used with the fiber optic elements being coupled to respective fuel nozzles and either to the photodetector or, wherein a plurality of photodetectors are used, to respective ones of the plurality of photodetectors. The signal processor can include a digital signal processor.

Local Burn-Up Effects in the NBSR Fuel Element

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

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

DOEpatents

Draley, J.E.; Ruther, W.E.

1959-01-27

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Crash Simulation of a Vertical Drop Test of a Commuter-Class Aircraft

NASA Technical Reports Server (NTRS)

Jackson, Karen E.; Fasanella, Edwin L.

2004-01-01

A finite element model of an ATR42-300 commuter-class aircraft was developed and a crash simulation was executed. Analytical predictions were correlated with data obtained from a 30-ft/s (9.14-m/s) vertical drop test of the aircraft. The purpose of the test was to evaluate the structural response of the aircraft when subjected to a severe, but survivable, impact. The aircraft was configured with seats, dummies, luggage, and other ballast. The wings were filled with 8,700 lb. (3,946 kg) of water to represent the fuel. The finite element model, which consisted of 57,643 nodes and 62,979 elements, was developed from direct measurements of the airframe geometry. The seats, dummies, luggage, fuel, and other ballast were represented using concentrated masses. The model was executed in LS-DYNA, a commercial code for performing explicit transient dynamic simulations. Predictions of structural deformation and selected time-history responses were generated. The simulation was successfully validated through extensive test-analysis correlation.

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.

Thickness effects of yttria-doped ceria interlayers on solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Fan, Zeng; An, Jihwan; Iancu, Andrei; Prinz, Fritz B.

2012-11-01

Determining the optimal thickness range of the interlayed yttria-doped ceria (YDC) films promises to further enhance the performance of solid oxide fuel cells (SOFCs) at low operating temperatures. The YDC interlayers are fabricated by the atomic layer deposition (ALD) method with one super cycle of the YDC deposition consisting of 6 ceria deposition cycles and one yttria deposition cycle. YDC films of various numbers of ALD super cycles, ranging from 2 to 35, are interlayered into bulk fuel cells with a 200 um thick yttria-stabilized zirconia (YSZ) electrolyte. Measurements and analysis of the linear sweep voltammetry of these fuel cells reveal that the performance of the given cells is maximized at 10 super cycles. Auger elemental mapping and X-ray photoelectron spectroscopy (XPS) techniques are employed to determine the film completeness, and they verify 10 super cycles of YDC to be the critical thickness point. This optimal YDC interlayer condition (6Ce1Y × 10 super cycles) is applied to the case of micro fuel cells as well, and the average performance enhancement factor is 1.4 at operating temperatures of 400 and 450 °C. A power density of 1.04 W cm-2 at 500 °C is also achieved with the optimal YDC recipe.

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

DOEpatents

McCorkle, W.H.

1960-02-23

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

Chemically Modified Metal Oxide Nanostructure for Photoelectrochemical Water Splitting

NASA Astrophysics Data System (ADS)

Wang, Gongming

Hydrogen gas is chemical fuel with high energy density, and represents a clean, renewable and carbon-free burning fuel, which has the potential to solve the more and more urgent energy crisis in today's society. Inspired by natural photosynthesis, artificial photosynthesis to generate hydrogen energy has attracted a lot of attentions in the field of chemistry, physics and material. Photoelectrochemical water splitting based on semiconductors represents a green and low cost method to generate hydrogen fuel. However, the current overall efficiency of solar to hydrogen is quite low, due to some intrinsic limitations such as bandgap, diffusion distance, carrier lifetime and photostability of semiconductors. Although nanostructured semiconductors can improve their photoelectrochemical water splitting performance to some extent, by increasing electrolyte accessible area and shortening minority carrier diffusion distance, nanostructure engineering cannot change their intrinsic electronic properties. Recent development in chemically modified nanostructures such as surface catalyst decoration, element doping, plasmonic modification and interfacial hetero-junction design have led to significant advancement in the photoelectrochemical water splitting, by improving surface reaction kinetics and charge separation, transportation and collection efficiency. In this thesis, I will give a detailed discussion on the chemically modified metal oxide nanostructures for photoelectrocemical hydrogen generation, with a focus on the element doping, hydrogen treatment and catalyst modification. I have demonstrated nitrogen doping on ZnO and Ti doping on hematite can improve their photoelectrochemical performance. In addition, we found hydrogen treatment is a general and effective method to improve the photocatalytic performance, by increasing their carrier desities. Hydrogen treatment has been demonstrated on TiO2, WO3 and BiVO4. In the end, we also used electrochemical catalyt to modify these metal oxide photoelectrode for waste water treatment and chemical fuel generation.

Simulation and in situ measurement of stress distribution in a polymer electrolyte membrane fuel cell stack

NASA Astrophysics Data System (ADS)

de la Cruz, Javier; Cano, Ulises; Romero, Tatiana

2016-10-01

A critical parameter for PEM fuel cell's electric contact is the nominal clamping pressure. Predicting the mechanical behavior of all components in a fuel cell stack is a very complex task due to the diversity of materials properties. Prior to the integration of a 3 kW PEMFC power plant, a numerical simulation was performed in order to obtain the mechanical stress distribution for two of the most pressure sensitive components of the stack: the membrane, and the graphite plates. The stress distribution of the above mentioned components was numerically simulated by finite element analysis and the stress magnitude for the membrane was confirmed using pressure films. Stress values were found within the elastic zone which guarantees mechanical integrity of fuel cell components. These low stress levels particularly for the membrane will allow prolonging the life and integrity of the fuel cell stack according to its design specifications.

A Multi-Methods Approach to HRA and Human Performance Modeling: A Field Assessment

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

Jacques Hugo; David I Gertman

2012-06-01

The Advanced Test Reactor (ATR) is a research reactor at the Idaho National Laboratory is primarily designed and used to test materials to be used in other, larger-scale and prototype reactors. The reactor offers various specialized systems and allows certain experiments to be run at their own temperature and pressure. The ATR Canal temporarily stores completed experiments and used fuel. It also has facilities to conduct underwater operations such as experiment examination or removal. In reviewing the ATR safety basis, a number of concerns were identified involving the ATR canal. A brief study identified ergonomic issues involving the manual handlingmore » of fuel elements in the canal that may increase the probability of human error and possible unwanted acute physical outcomes to the operator. In response to this concern, that refined the previous HRA scoping analysis by determining the probability of the inadvertent exposure of a fuel element to the air during fuel movement and inspection was conducted. The HRA analysis employed the SPAR-H method and was supplemented by information gained from a detailed analysis of the fuel inspection and transfer tasks. This latter analysis included ergonomics, work cycles, task duration, and workload imposed by tool and workplace characteristics, personal protective clothing, and operational practices that have the potential to increase physical and mental workload. Part of this analysis consisted of NASA-TLX analyses, combined with operational sequence analysis, computational human performance analysis (CHPA), and 3D graphical modeling to determine task failures and precursors to such failures that have safety implications. Experience in applying multiple analysis techniques in support of HRA methods is discussed.« less

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

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

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Top Ten Reasons for DEOX as a Front End to Pyroprocessing

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

B.R. Westphal; K.J. Bateman; S.D. Herrmann

A front end step is being considered to augment chopping during the treatment of spent oxide fuel by pyroprocessing. The front end step, termed DEOX for its emphasis on decladding via oxidation, employs high temperatures to promote the oxidation of UO2 to U3O8 via an oxygen carrier gas. During oxidation, the spent fuel experiences a 30% increase in lattice structure volume resulting in the separation of fuel from cladding with a reduced particle size. A potential added benefit of DEOX is the removal of fission products, either via direct release from the broken fuel structure or via oxidation and volatilizationmore » by the high temperature process. Fuel element chopping is the baseline operation to prepare spent oxide fuel for an electrolytic reduction step. Typical chopping lengths range from 1 to 5 mm for both individual elements and entire assemblies. During electrolytic reduction, uranium oxide is reduced to metallic uranium via a lithium molten salt. An electrorefining step is then performed to separate a majority of the fission products from the recoverable uranium. Although DEOX is based on a low temperature oxidation cycle near 500oC, additional conditions have been tested to distinguish their effects on the process.[1] Both oxygen and air have been utilized during the oxidation portion followed by vacuum conditions to temperatures as high as 1200oC. In addition, the effects of cladding on fission product removal have also been investigated with released fuel to temperatures greater than 500oC.« less

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.

PATHFINDER ATOMIC POWER PLANT TECHNICAL PROGRESS REPORT FOR JULY 1, 1959- SEPTEMBER 30, 1959

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

None

1960-10-31

ABS>Fuel Element Research and Development. Dynamic and static corrosion tests on 8001 Al were completed. Annealmmmg of 1100 cladding on 5083 and M400 cladding on X2219 were tested at 500 deg C, and investigation continued on producing X8101 Al alloy cladding in tube plates by extrusion. Boiler fuel element capsule irradiation tests and subassembly tests are described Heat transfer loop studies and fuel fabrication for the critical facility are reported. Boiler fuel element mechanical design and testing progress is desc ribed. and the superheater fuel element temperature evaluating routine is discussed. Low- enrichment superheater fuel element development included design studiesmore » and stainless steel powder and UO/sub 2/ powder fabrication studies Reactor Mechanical Studies. Research is reported on vessel and structure design, fabrication, and testing, recirculation system design, steam separator tests, and control rod studies. Nuclear Analysis. Reactor physics studies are reported on nuclear constants, baffle plate analysis, comparison of core representations, delayed neutron fraction. and shielding analysis of the reactor building. Reactor and system dynamics and critical experiments were also studied. Chemistry. Progress is reported on recombiner. radioactive gas removal and storage, ion exchanger and radiochemical processing. (For preceding period see ACNP-5915.) (T.R.H.)« 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.

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

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Study for the optimization of a transport aircraft wing for maximum fuel efficiency. Volume 1: Methodology, criteria, aeroelastic model definition and results

NASA Technical Reports Server (NTRS)

Radovcich, N. A.; Dreim, D.; Okeefe, D. A.; Linner, L.; Pathak, S. K.; Reaser, J. S.; Richardson, D.; Sweers, J.; Conner, F.

1985-01-01

Work performed in the design of a transport aircraft wing for maximum fuel efficiency is documented with emphasis on design criteria, design methodology, and three design configurations. The design database includes complete finite element model description, sizing data, geometry data, loads data, and inertial data. A design process which satisfies the economics and practical aspects of a real design is illustrated. The cooperative study relationship between the contractor and NASA during the course of the contract is also discussed.

Affordable Development and Demonstration of a Small NTR Engine and Stage: How Small is Big Enough?

NASA Technical Reports Server (NTRS)

Borowski, S. K.; Sefcik, R. J.; Fittje, J. E.; McCurdy, D. R.; Qualls, A. L.; Schnitzler, B. G.; Werner, J.; Weitzberg, A.; Joyner, C. R.

2015-01-01

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations The ongoing NTP project, funded by NASAs Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities:(1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options;(2) Engine conceptual design;(3) Mission analysis and engine requirements definition;(4) Identification of affordable options for ground testing; and(5) Formulation of an affordable and sustainable NTP development program Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klbf-class engine were developed during FYs 13-14 using heritage fuel element designs for both RoverNERVA Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down selection between GC and cermet fuel was required An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE plan was produced by GRC, DOE industry for the AES program. Assumptions, considerations and key task activities are presented here Two small (7.5 and 16.5 klbf) engine sizes were considered for ground and flight technology demonstration within a 10-year timeframe; their ability to support future human exploration missions was also examined and a recommendation on a preferred size is provided.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

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

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes

2013-10-01

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column weremore » observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.« less

Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

DOEpatents

Cowell, G.K.; DiGuiseppe, C.P.

1982-10-28

A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

Investigations of Multiple Swirl-Venturi Fuel Injector Concepts: Recent Experimental Optical Measurement Results for 1-Point, 7-Point, and 9-Point Configurations

NASA Technical Reports Server (NTRS)

Hicks, Yolanda R.; Anderson, Robert C.; Tedder, Sarah A.; Tacina, Kathleen M.

2015-01-01

This paper presents results obtained during testing in optically-accessible, JP8-fueled, flame tube combustors using swirl-venturi lean direct injection (LDI) research hardware. The baseline LDI geometry has 9 fuel/air mixers arranged in a 3 x 3 array within a square chamber. 2-D results from this 9-element array are compared to results obtained in a cylindrical combustor using a 7-element array and a single element. In each case, the baseline element size remains the same. The effect of air swirler angle, and element arrangement on the presence of a central recirculation zone are presented. Only the highest swirl number air swirler produced a central recirculation zone for the single element swirl-venturi LDI and the 9-element LDI, but that same swirler did not produce a central recirculation zone for the 7-element LDI, possibly because of strong interactions due to element spacing within the array.

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

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

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Aluminum hydroxide coating thickness measurements and brushing tests on K West Basin fuel elements

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

Pitner, A.L.

1998-09-11

Aluminum hydroxide coating thicknesses were measured on fuel elements stored in aluminum canisters in K West Basin using specially developed eddy current probes . The results were used to estimate coating inventories for MCO fuel,loading. Brushing tests successfully demonstrated the ability to remove the coating if deemed necessary prior to MCO loading.

Photographic combustion characterization of LOX/hydrocarbon type propellants

NASA Technical Reports Server (NTRS)

Judd, D. C.

1979-01-01

Single element injectors and two fuels were tested with the aim of photographically characterizing observed combustion phenomena. The three injectors tested were the O-F-O triplet, the transverse like on like (TLOL), and the rectangular unlike doublet (RUD). The fuels tested were RP-1 and propane. The hot firings were conducted in a specifically constructed chamber fitted with quartz windows for photographically viewing the impingement spray field. All LOX/HC testing demonstrated coking with the RP-1 fuel leaving far more soot than the propane fuel. No fuel freezing or popping was experienced under the test conditions evaluated. Carbon particle emission and combustion light brilliance increased with Pc for both fuels although RP-1 was far more energetic in this respect. The RSS phenomena appear to be present in the high Pc tests as evidenced by striations in the spray pattern and by separate fuel rich and oxidizer rich areas. The RUD element was also tested as a fuel rich gas generator element by switching the propellant circuits. Excessive sooting occurred at this low mixture ratio (0.55), precluding photographic data.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

ELECTROLYTIC SEPARATION PROCESS AND APPARATUS

DOEpatents

McLain, M.E. Jr.; Roberts, M.W.

1962-03-01

A method is given for dissolving stainless steel-c lad fuel elements in dilute acids such as half normal sulfuric acid. The fuel element is made the anode in a Y-shaped electrolytic cell which has a flowing mercury cathode; the stainless steel elements are entrained in the mercury and stripped therefrom by a continuous process. (AEC)

Nuclear breeder reactor fuel element with silicon carbide getter

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1987-01-01

An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

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

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

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

2015-06-15

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

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

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

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

Fuel Pin Behavior Under the Slow Power Ramp Transients in the CABRI-2 Experiments

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

Charpenel, Jean; Lemoine, Francette; Sato, Ikken

Slow ramp-type transient-overpower tests were performed within the framework of the international CABRI-2 experimental program. The implemented power transients of {approx}1% nominal power/s correspond to a control rod withdrawal-type accident in a liquid-metal-cooled fast breeder reactor (FBR). The analysis of the tests includes the information elements derived from the hodoscope signals, which were assessed quantitatively and supported by destructive and nondestructive posttest examinations. These tests, performed with fuels of various geometries, demonstrated the high margin to failure of such FBR fuel pins within the expected power level before the emergency reactor shutdown. At the same time, these tests performed withmore » high- and low-smear-density industrial pins led to clarification of the influence of pellet design on fuel pin behavior under high overpower condition. With the high-smear-density solid fuel pellet pin of high burnup level, the retained gaseous fission products played an important role in the solid fuel swelling, leading to clad deformation and failure at a maximum heating rate of 81 kW.m{sup -1}, which is much greater than the end-of-life (EOL) linear rating of the pin. With the low smear-density annular pellet pin, an important fuel swelling takes place, leading to degradation of the fuel thermal conductivity. This effect was detected at the power level around 73 kW.m{sup -1}, which is also much higher than the EOL value of the pin. Furthermore, the absence of clad deformation, and consequently of failure even at the power level going up to 134.7 kW.m{sup -1}, confirmed the very high margin to failure. In consequence, it was clarified that gaseous fission products have significant effects on failure threshold as well as on thermal performance during overpower condition, and such effects are significantly dependent on fuel design and power operation conditions.« less

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

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

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

Gaseous swelling of U 3 Si 2 during steady-state LWR operation: A rate theory investigation

DOE PAGES

Miao, Yinbin; Gamble, Kyle A.; Andersson, David; ...

2017-07-25

Rate theory simulations of fission gas behavior in U 3Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. We developed a model of U 3Si 2 and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3Si 2 temperature is expected to be below 1000 K,more » intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.« less

Fuel containment and damage tolerance in large composite primary aircraft structures. Phase 2: Testing

NASA Technical Reports Server (NTRS)

Sandifer, J. P.; Denny, A.; Wood, M. A.

1985-01-01

Technical issues associated with fuel containment and damage tolerance of composite wing structures for transport aircraft were investigated. Material evaluation tests were conducted on two toughened resin composites: Celion/HX1504 and Celion/5245. These consisted of impact, tension, compression, edge delamination, and double cantilever beam tests. Another test series was conducted on graphite/epoxy box beams simulating a wing cover to spar cap joint configuration of a pressurized fuel tank. These tests evaluated the effectiveness of sealing methods with various fastener types and spacings under fatigue loading and with pressurized fuel. Another test series evaluated the ability of the selected coatings, film, and materials to prevent fuel leakage through 32-ply AS4/2220-1 laminates at various impact energy levels. To verify the structural integrity of the technology demonstration article structural details, tests were conducted on blade stiffened panels and sections. Compression tests were performed on undamaged and impacted stiffened AS4/2220-1 panels and smaller element tests to evaluate stiffener pull-off, side load and failsafe properties. Compression tests were also performed on panels subjected to Zone 2 lightning strikes. All of these data were integrated into a demonstration article representing a moderately loaded area of a transport wing. This test combined lightning strike, pressurized fuel, impact, impact repair, fatigue and residual strength.

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Liquid fuel injection elements for rocket engines

NASA Technical Reports Server (NTRS)

Cox, George B., Jr. (Inventor)

1993-01-01

Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

FLUID MODERATED REACTOR

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

High performance fuel element with end seal

DOEpatents

Lee, Gary E.; Zogg, Gordon J.

1987-01-01

A nuclear fuel element comprising an elongate block of refractory material having a generally regular polygonal cross section. The block includes parallel, spaced, first and second end surfaces. The first end surface has a peripheral sealing flange formed thereon while the second end surface has a peripheral sealing recess sized to receive the flange. A plurality of longitudinal first coolant passages are positioned inwardly of the flange and recess. Elongate fuel holes are separate from the coolant passages and disposed inwardly of the flange and the recess. The block is further provided with a plurality of peripheral second coolant passages in general alignment with the flange and the recess for flowing coolant. The block also includes two bypasses for each second passage. One bypass intersects the second passage adjacent to but spaced from the first end surface and intersects a first passage, while the other bypass intersects the second passage adjacent to but spaced from the second end surface and intersects a first passage so that coolant flowing through the second passages enters and exits the block through the associated first passages.

Finite element assisted prediction of ductile fracture in sheet bulging

NASA Astrophysics Data System (ADS)

Donald, Bryan J. Mac; Lorza, Ruben Lostado; Yoshihara, Shoichiro

2017-10-01

With growing demand for energy efficiency, there is much focus on reducing oil consumption rates and utilising alternative fuels. A contributor to the solution in this area is to produce lighter vehicles that are more fuel efficient and/or allow for the use of alternative fuel sources (e.g. electric powered automobiles). Near-net-shape manufacturing processes such as hydroforming have great potential to reduce structural weight while still maintaining structural strength and performance. Finite element analysis techniques have proved invaluable in optimizing such hydroforming processes, however, the majority of such studies have used simple predictors of failure which are usually yield criteria such as von Mises stress. There is clearly potential to obtain more optimal solutions using more advanced predictors of failure. This paper compared the Von Mises stress failure criteria and the Oyane's ductile fracture criteria in the sheet hydroforming of magnesium alloys. It was found that the results obtained from the models which used Oyane's ductile fracture criteria were more realistic than those obtained from those that used Von Mises stress as a failure criteria.

Structural integrity of a confinement vessel for testing nuclear fuels for space propulsion

NASA Astrophysics Data System (ADS)

Bergmann, V. L.

Nuclear propulsion systems for rockets could significantly reduce the travel time to distant destinations in space. However, long before such a concept can become reality, a significant effort must be invested in analysis and ground testing to guide the development of nuclear fuels. Any testing in support of development of nuclear fuels for space propulsion must be safely contained to prevent the release of radioactive materials. This paper describes analyses performed to assess the structural integrity of a test confinement vessel. The confinement structure, a stainless steel pressure vessel with bolted flanges, was designed for operating static pressures in accordance with the ASME Boiler and Pressure Vessel Code. In addition to the static operating pressures, the confinement barrier must withstand static overpressures from off-normal conditions without releasing radioactive material. Results from axisymmetric finite element analyses are used to evaluate the response of the confinement structure under design and accident conditions. For the static design conditions, the stresses computed from the ASME code are compared with the stresses computed by the finite element method.

Improved gas tagging and cover gas combination for nuclear reactor

DOEpatents

Gross, K.C.; Laug, M.T.

1983-09-26

The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Foote, F.G.; Jette, E.R.

1963-05-01

A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

Method and apparatus for diagnosing breached fuel elements

DOEpatents

Gross, K.C.; Lambert, J.D.B.; Nomura, S.

1987-03-02

The invention provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor. A detection system measures the activity of isotopes from the cover gas in the reactor. A data acquisition and processing system monitors the detection system and corrects for the effects of the cover-gas clean up system on the measured activity and further calculates the derivative curve of the corrected activity as a function of time. A plotting system graphs the derivative curve, which represents the instantaneous release rate of fission gas from a breached fuel element. 8 figs.

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

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

1959-05-12

Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

A feasibility study of reactor-based deep-burn concepts.

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

Kim, T. K.; Taiwo, T. A.; Hill, R. N.

2005-09-16

A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team.more » The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling or by operating with lower material temperatures (colder core). Results also showed that the transmutation performance of the one-pass deep-burn concept is sensitive to the initial TRU vector, primarily because longer cooling time reduces the fissile content (Pu-241 specifically.) With a cooling time of 5 years, the TRU consumption increases to 67%, while conversely, with 20-year cooling the TRU consumption is about 58%. For the two-pass DB-MHR (TRU recycling option), a fuel packing fraction of about 30% is required in the second pass (the recycled TRU). It was found that using a heterogeneous core (homogeneous fuel element) concept, the TRU consumption is dependent on the cooling interval before the 2nd pass, again due to Pu-241 decay during the time lag between the first pass fuel discharge and the second pass fuel charge. With a cooling interval of 7 years (5 and 2 years before and after reprocessing) a TRU consumption of 55% is obtained. With an assumed ''no cooling'' interval, the TRU consumption is 63%. By using a cylindrical core to reduce neutron leakage, TRU consumption of the case with 7-year cooling interval increases to 58%. For a two-pass concept using a heterogeneous fuel element (and homogeneous core) with first and second pass volume ratio of 2:1, the TRU consumption is 62.4%. Finally, the repository loading benefits arising from the deep-burn and Inert Matrix Fuel (IMF) concepts were estimated and compared, for the same initial TRU vector. The DB-MHR concept resulted in slightly higher TRU consumption and repository loading benefit compared to the IMF concept (58.1% versus 55.1% for TRU consumption and 2.0 versus 1.6 for estimated repository loading benefit).« less

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

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

Feldman, E.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodatemore » an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well.« less

Faecal-wood biomass co-combustion and ash composition analysis.

PubMed

Somorin, Tosin Onabanjo; Kolios, Athanasios J; Parker, Alison; McAdam, Ewan; Williams, Leon; Tyrrel, Sean

2017-09-01

Fuel blending is a widely used approach in biomass combustion, particularly for feedstocks with low calorific value and high moisture content. In on-site sanitation technologies, fuel blending is proposed as a pre-treatment requirement to reduce moisture levels and improve the physiochemical properties of raw faeces prior to drying. This study investigates the co-combustion performance of wood dust: raw human faeces blends at varying air-to-fuel ratios in a bench-scale combustor test rig. It concludes with ash composition analyses and discusses their potential application and related problems. The study shows that a 50:50 wood dust (WD): raw human faeces (FC) can reduce moisture levels in raw human faeces by ∼40% prior to drying. The minimum acceptable blend for treating moist faeces without prior drying at a combustion air flow rate of 14-18 L/min is 30:70 WD: FC. For self-sustained ignition and flame propagation, the minimum combustion temperature required for conversion of the fuel to ash is ∼400 °C. The most abundant elements in faecal ash are potassium and calcium, while elements such as nickel, aluminium and iron are in trace quantities. This suggests the potential use of faecal ash as a soil conditioner, but increases the tendency for fly ash formation and sintering problems.

FLUSH - PREDICTION OF FLOW PARAMETERS OF SLUSH HYDROGEN

NASA Technical Reports Server (NTRS)

Hardy, T.

1994-01-01

Slush hydrogen, a mixture of the solid and liquid phases of hydrogen, is a possible source of fuel for the National Aerospace Plane (NASP) Project. Advantages of slush hydrogen over liquid hydrogen include greater heat capacity and greater density. However, practical use of slush hydrogen as a fuel requires systems of lines, valves, etc. which are designed to deliver the fuel in slush form with minimal solid loss as a result of pipe heating or flow friction. Engineers involved with the NASP Project developed FLUSH to calculate the pressure drop and slush hydrogen solid fraction loss for steady-state, one-dimensional flow. FLUSH solves the steady-state, one-dimensional energy equation and the Bernoulli equation for pipe flow. The program performs these calculations for each two-node element--straight pipe length, elbow, valve, fitting, or other part of the piping system--specified by the user. The user provides flow rate, upstream pressure, initial solid hydrogen fraction, element heat leak, and element parameters such as length and diameter. For each element, FLUSH first calculates the pressure drop, then figures the slush solid fraction exiting the element. The code employs GASPLUS routines to calculate thermodynamic properties for the slush hydrogen. FLUSH is written in FORTRAN IV for DEC VAX series computers running VMS. An executable is provided on the tape. The GASPLUS physical properties routines which are required for building the executable are included as one object library on the program media (full source code for GASPLUS is available separately as COSMIC Program Number LEW-15091). FLUSH is available in DEC VAX BACKUP format on a 9-track 1600 BPI magnetic tape (standard media) or on a TK50 tape cartridge. FLUSH was developed in 1989.

FUEL ELEMENT CONSTRUCTION

DOEpatents

Zumwalt, L.R.

1961-08-01

Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

DOEpatents

Brooks, H.

1960-04-26

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

KINETICS OF TREAT USED AS A TEST REACTOR

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

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

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

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Ion chromatographic determination of sulfur in fuels

NASA Technical Reports Server (NTRS)

Mizisin, C. S.; Kuivinen, D. E.; Otterson, D. A.

1978-01-01

The sulfur content of fuels was determined using an ion chromatograph to measure the sulfate produced by a modified Parr bomb oxidation. Standard Reference Materials from the National Bureau of Standards, of approximately 0.2 + or - 0.004% sulfur, were analyzed resulting in a standard deviation no greater than 0.008. The ion chromatographic method can be applied to conventional fuels as well as shale-oil derived fuels. Other acid forming elements, such as fluorine, chlorine and nitrogen could be determined at the same time, provided that these elements have reached a suitable ionic state during the oxidation of the fuel.

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

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

Rest, J.; Hofman, G.L.

1999-04-01

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

A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

NASA Astrophysics Data System (ADS)

Navarro, Jorge

The goal of this study presented is to determine the best available nondestructive technique necessary to collect validation data as well as to determine burnup and cooling time of the fuel elements on-site at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal, the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements nondestructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed were used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results, it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however, in order to enhance the quality of the spectra collected using this scintillator, a deconvolution method was developed. Following the development of the deconvolution method for ATR applications, the technique was tested using one-isotope, multi-isotope, and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr 3 detector in an above the water configuration and deconvolution algorithms.

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: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

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

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

High density dispersion fuel

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

Hofman, G.L.

1996-09-01

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

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

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

Hakan Ozaltun; Pavel Medvedev

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate frommore » RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.« less

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

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.

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

NASA Technical Reports Server (NTRS)

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

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 neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a 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 can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Radiological Characterization Methodology of INEEL Stored RH-TRU Waste from ANL-E

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

Rajiv N. Bhatt

2003-02-01

An Acceptable Knowledge (AK)-based radiological characterization methodology is being developed for RH TRU waste generated from ANL-E hot cell operations performed on fuel elements irradiated in the EBR-II reactor. The methodology relies on AK for composition of the fresh fuel elements, their irradiation history, and the waste generation and collection processes. Radiological characterization of the waste involves the estimates of the quantities of significant fission products and transuranic isotopes in the waste. Methods based on reactor and physics principles are used to achieve these estimates. Because of the availability of AK and the robustness of the calculation methods, the AK-basedmore » characterization methodology offers a superior alternative to traditional waste assay techniques. Using this methodology, it is shown that the radiological parameters of a test batch of ANL-E waste is well within the proposed WIPP Waste Acceptance Criteria limits.« less

Radiological Characterization Methodology for INEEL-Stored Remote-Handled Transuranic (RH TRU) Waste from Argonne National Laboratory-East

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

Kuan, P.; Bhatt, R.N.

2003-01-14

An Acceptable Knowledge (AK)-based radiological characterization methodology is being developed for RH TRU waste generated from ANL-E hot cell operations performed on fuel elements irradiated in the EBR-II reactor. The methodology relies on AK for composition of the fresh fuel elements, their irradiation history, and the waste generation and collection processes. Radiological characterization of the waste involves the estimates of the quantities of significant fission products and transuranic isotopes in the waste. Methods based on reactor and physics principles are used to achieve these estimates. Because of the availability of AK and the robustness of the calculation methods, the AK-basedmore » characterization methodology offers a superior alternative to traditional waste assay techniques. Using the methodology, it is shown that the radiological parameters of a test batch of ANL-E waste is well within the proposed WIPP Waste Acceptance Criteria limits.« less

Upgraded HFIR Fuel Element Welding System

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

Sease, John D

2010-02-01

The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. Inmore » recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.« less

Apollo 12 Mission image - Alan Bean unloads ALSEP RTG fuel element

NASA Image and Video Library

1969-11-19

AS12-46-6790 (19 Nov. 1969) --- Astronaut Alan L. Bean, lunar module pilot, is photographed at quadrant II of the Lunar Module (LM) during the first Apollo 12 extravehicular activity (EVA) on the moon. This picture was taken by astronaut Charles Conrad Jr., commander. Here, Bean is using a fuel transfer tool to remove the fuel element from the fuel cask mounted on the LM's descent stage. The fuel element was then placed in the Radioisotope Thermoelectric Generator (RTG), the power source for the Apollo Lunar Surface Experiments Package (ALSEP) which was deployed on the moon by the two astronauts. The RTG is next to Bean's right leg. While astronauts Conrad and Bean descended in the LM "Intrepid" to explore the Ocean of Storms region of the moon, astronaut Richard F. Gordon Jr., command module pilot, remained with the Command and Service Modules (CSM) "Yankee Clipper" in lunar orbit.

Reliability analysis of dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Method for recovering catalytic elements from fuel cell membrane electrode assemblies

DOEpatents

Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

2012-06-26

A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

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

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T.; James P. Meagher; Prasad Apte

2002-12-31

This topical report summarizes work accomplished for the Program from November 1, 2001 to December 31, 2002 in the following task areas: Task 1: Materials Development; Task 2: Composite Development; Task 4: Reactor Design and Process Optimization; Task 8: Fuels and Engine Testing; 8.1 International Diesel Engine Program; 8.2 Nuvera Fuel Cell Program; and Task 10: Program Management. Major progress has been made towards developing high temperature, high performance, robust, oxygen transport elements. In addition, a novel reactor design has been proposed that co-produces hydrogen, lowers cost and improves system operability. Fuel and engine testing is progressing well, but wasmore » delayed somewhat due to the hiatus in program funding in 2002. The Nuvera fuel cell portion of the program was completed on schedule and delivered promising results regarding low emission fuels for transportation fuel cells. The evaluation of ultra-clean diesel fuels continues in single cylinder (SCTE) and multiple cylinder (MCTE) test rigs at International Truck and Engine. FT diesel and a BP oxygenate showed significant emissions reductions in comparison to baseline petroleum diesel fuels. Overall through the end of 2002 the program remains under budget, but behind schedule in some areas.« less

Combined catalysts for the combustion of fuel in gas turbines

DOEpatents

Anoshkina, Elvira V.; Laster, Walter R.

2012-11-13

A catalytic oxidation module for a catalytic combustor of a gas turbine engine is provided. The catalytic oxidation module comprises a plurality of spaced apart catalytic elements for receiving a fuel-air mixture over a surface of the catalytic elements. The plurality of catalytic elements includes at least one primary catalytic element comprising a monometallic catalyst and secondary catalytic elements adjacent the primary catalytic element comprising a multi-component catalyst. Ignition of the monometallic catalyst of the primary catalytic element is effective to rapidly increase a temperature within the catalytic oxidation module to a degree sufficient to ignite the multi-component catalyst.

Limits of thermochemical and photochemical syntheses of gaseous fuels: a finite-time thermodynamic analysis. Annual report, September 1983-February, 1985

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

Berry, R.S.

The objectives of this project are to develop methods for the evaluation of syntheses of gaseous fuels in terms of their optimum possible performance, particularly when they are required to supply those fuels at nonzero rates. The first objective is entirely in the tradition of classical thermodynamics, the processes, given the characteristics and constraints that define them. The new element which this project introduces is the capability to set limits more realistic than those from classical thermodynamics, by the inclusion of the influence of the rate or duration of a process on its performance. The development of these analyses ismore » a natural step in the evolution represented by the evaluative papers of Appendix IV, e.g., by Funk et al., Abraham, Shinnar, Bilgen and Fletcher. A second objective is to determine how any given process should be carried out, within its constraints, in order to yield its optimum performance and to use this information whenever possible to help guide the design of that process.« less

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

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

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Refinery Integration of By-Products from Coal-Derived Jet Fuels

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

Caroline Clifford; Andre Boehman; Chunshan Song

2008-03-31

The final report summarizes the accomplishments toward project goals during length of the project. The goal of this project was to integrate coal into a refinery in order to produce coal-based jet fuel, with the major goal to examine the products other than jet fuel. These products are in the gasoline, diesel and fuel oil range and result from coal-based jet fuel production from an Air Force funded program. The main goal of Task 1 was the production of coal-based jet fuel and other products that would need to be utilized in other fuels or for non-fuel sources, using knownmore » refining technology. The gasoline, diesel fuel, and fuel oil were tested in other aspects of the project. Light cycle oil (LCO) and refined chemical oil (RCO) were blended, hydrotreated to removed sulfur, and hydrogenated, then fractionated in the original production of jet fuel. Two main approaches, taken during the project period, varied where the fractionation took place, in order to preserve the life of catalysts used, which includes (1) fractionation of the hydrotreated blend to remove sulfur and nitrogen, followed by a hydrogenation step of the lighter fraction, and (2) fractionation of the LCO and RCO before any hydrotreatment. Task 2 involved assessment of the impact of refinery integration of JP-900 production on gasoline and diesel fuel. Fuel properties, ignition characteristics and engine combustion of model fuels and fuel samples from pilot-scale production runs were characterized. The model fuels used to represent the coal-based fuel streams were blended into full-boiling range fuels to simulate the mixing of fuel streams within the refinery to create potential 'finished' fuels. The representative compounds of the coal-based gasoline were cyclohexane and methyl cyclohexane, and for the coal-base diesel fuel they were fluorine and phenanthrene. Both the octane number (ON) of the coal-based gasoline and the cetane number (CN) of the coal-based diesel were low, relative to commercial fuels ({approx}60 ON for coal-based gasoline and {approx}20 CN for coal-based diesel fuel). Therefore, the allowable range of blending levels was studied where the blend would achieve acceptable performance. However, in both cases of the coal-based fuels, their ignition characteristics may make them ideal fuels for advanced combustion strategies where lower ON and CN are desirable. Task 3 was designed to develop new approaches for producing ultra clean fuels and value-added chemicals from refinery streams involving coal as a part of the feedstock. It consisted of the following three parts: (1) desulfurization and denitrogenation which involves both new adsorption approach for selective removal of nitrogen and sulfur and new catalysts for more effective hydrotreating and the combination of adsorption denitrogenation with hydrodesulfurization; (2) saturation of two-ring aromatics that included new design of sulfur resistant noble-metal catalysts for hydrogenation of naphthalene and tetralin in middle distillate fuels, and (3) value-added chemicals from naphthalene and biphenyl, which aimed at developing value-added organic chemicals from refinery streams such as 2,6-dimethylnaphthalene and 4,4{prime}-dimethylbiphenyl as precursors to advanced polymer materials. Major advances were achieved in this project in designing the catalysts and sorbent materials, and in developing fundamental understanding. The objective of Task 4 was to evaluate the effect of introducing coal into an existing petroleum refinery on the fuel oil product, specifically trace element emissions. Activities performed to accomplish this objective included analyzing two petroleum-based commercial heavy fuel oils (i.e., No. 6 fuel oils) as baseline fuels and three co-processed fuel oils, characterizing the atomization performance of a No. 6 fuel oil, measuring the combustion performance and emissions of the five fuels, specifically major, minor, and trace elements when fired in a watertube boiler designed for natural gas/fuel oil, and determining the boiler performance when firing the five fuels. Two different co-processed fuel oils were tested: one that had been partially hydrotreated, and the other a product of fractionation before hydrotreating. Task 5 focused on examining refining methods that would utilize coal and produce thermally stable jet fuel, included delayed coking and solvent extraction. Delayed coking was done on blends of decant oil and coal, with the goal to produce a premium carbon product and liquid fuels. Coking was done on bench scale and large laboratory scale cokers. Two coals were examined for co-coking, using Pittsburgh seam coal and Marfork coal product. Reactions in the large, laboratory scaled coker were reproducible in yields of products and in quality of products. While the co-coke produced from both coals was of sponge coke quality, minerals left in the coke made it unacceptable for use as anode or graphite grade filler.« less

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

A method for determining optimum phasing of a multiphase propulsion system for a single-stage vehicle with linearized inert weight

NASA Technical Reports Server (NTRS)

Martin, J. A.

1974-01-01

A general analytical treatment is presented of a single-stage vehicle with multiple propulsion phases. A closed-form solution for the cost and for the performance and a derivation of the optimal phasing of the propulsion are included. Linearized variations in the inert weight elements are included, and the function to be minimized can be selected. The derivation of optimal phasing results in a set of nonlinear algebraic equations for optimal fuel volumes, for which a solution method is outlined. Three specific example cases are analyzed: minimum gross lift-off weight, minimum inert weight, and a minimized general function for a two-phase vehicle. The results for the two-phase vehicle are applied to the dual-fuel rocket. Comparisons with single-fuel vehicles indicate that dual-fuel vehicles can have lower inert weight either by development of a dual-fuel engine or by parallel burning of separate engines from lift-off.

Thermal stress analysis of sulfur deactivated solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Zeng, Shumao; Parbey, Joseph; Yu, Guangsen; Xu, Min; Li, Tingshuai; Andersson, Martin

2018-03-01

Hydrogen sulfide in fuels can deactivate catalyst for solid oxide fuel cells, which has become one of the most critical challenges to stability. The reactions between sulfur and catalyst will cause phase changes, leading to increase in cell polarization and mechanical mismatch. A three-dimensional computational fluid dynamics (CFD) approach based on the finite element method (FEM) is thus used to investigate the polarization, temperature and thermal stress in a sulfur deactivated SOFC by coupling equations for gas-phase species, heat, momentum, ion and electron transport. The results indicate that sulfur in fuels can strongly affect the cell polarization and thermal stresses, which shows a sharp decrease in the vicinity of electrolyte when 10% nickel in the functional layer is poisoned, but they remain almost unchanged even when the poisoned Ni content was increased to 90%. This investigation is helpful to deeply understand the sulfur poisoning effects and also benefit the material design and optimization of electrode structure to enhance cell performance and lifetimes in various hydrocarbon fuels containing impurities.

NUCLEAR REACTOR CORE

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

Bell, F.R.

1963-02-01

A nuclear reactor core composed of a number of identical elements of solid moderator material fitted together was designed. Each moderator element is apertured to provide channels for fuel and coolant. The elements have an external shape which permits them to be stacked in layers with similar elements, with the surfaces of adjacent elements fitting and in contact with each other. The cross section of the element is of a general hexagonal shape with identations and protrusions, so that the elements can be fitted together. The described core should not be liable to fracture under transverse loading. Specific arrangements ofmore » moderator elements and fuel and coolant apertures are described. (M.P.G.)« less

Transmutation Scoping Studies for a Chloride Molten Salt Reactor

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

Heidet, Florent; Feng, Bo; Kim, Taek

2016-01-01

Over the past few years, there has been strong renewed interest from private industry, mostly from start-up enterprises, in molten salt reactor (MSR) technologies because of the unique properties of this class of reactors. These are reactors in which the fuel is homogeneously mixed with the coolant in the form of liquid salts and is circulated continuously into and out of the active core region with on-line fuel management, salt treatment, and salt processing. In response to such wide-spread interest, Argonne National Laboratory is expanding its well-established reactor modelling and simulation expertise and infrastructure to enable detailed analysis and designmore » of MSRs. The tools being developed are able to simulate the continuous fuel flow, the complex on-line fuel management and elemental removal processes (e.g., fission product removal) using depletion steps representative of a real MSR system. Leveraging these capabilities, a parametric study on the transmutation performance of a simplified actinide-burning MSR concept that uses a chloride-based salt was performed. This type of salt has attracted attention over the more commonly discussed fluoride-based salts since no tritium is produced as a result of irradiation and it is compatible with a fast neutron spectrum. The studies discussed in this paper examine the performance of a burner MSR design with a fixed core size and power density over a range of possible fuel salt molar ratios with NaCl-MgCl2 as the carrier salt. The intent is to quantify the impact on the required transuranics content of the make-up fuel, the actinide transmutation rates, and other performance characteristics for typical burner MSR designs.« less

Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

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

Ozaltun, Hakan; Medvedev, Pavel G

2015-06-01

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance havemore » been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.« less

Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

PubMed

Saqib, Naeem; Bäckström, Mattias

2014-12-01

Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-05-01

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

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.

Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

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

Rowsell, David Leon

This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

The manufacture of LEU fuel elements at Dounreay

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

Gibson, J.

1997-08-01

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

Corrosion protected, multi-layer fuel cell interface

DOEpatents

Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

1986-01-01

An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

A numerical investigation of the influence of radiation and moisture content on pyrolysis and ignition of a leaf-like fuel element

Treesearch

B.L. Yashwanth; B. Shotorban; S. Mahalingam; C.W. Lautenberger; David Weise

2016-01-01

The effects of thermal radiation and moisture content on the pyrolysis and gas phase ignition of a solid fuel element containing high moisture content were investigated using the coupled Gpyro3D/FDS models. The solid fuel has dimensions of a typical Arctostaphylos glandulosa leaf which is modeled as thin cellulose subjected to radiative heating on...

AST Critical Propulsion and Noise Reduction Technologies for Future Commercial Subsonic Engines Area of Interest 1.0: Reliable and Affordable Control Systems

NASA Technical Reports Server (NTRS)

Myers, William; Winter, Steve

2006-01-01

The General Electric Reliable and Affordable Controls effort under the NASA Advanced Subsonic Technology (AST) Program has designed, fabricated, and tested advanced controls hardware and software to reduce emissions and improve engine safety and reliability. The original effort consisted of four elements: 1) a Hydraulic Multiplexer; 2) Active Combustor Control; 3) a Variable Displacement Vane Pump (VDVP); and 4) Intelligent Engine Control. The VDVP and Intelligent Engine Control elements were cancelled due to funding constraints and are reported here only to the state they progressed. The Hydraulic Multiplexing element developed and tested a prototype which improves reliability by combining the functionality of up to 16 solenoids and servo-valves into one component with a single electrically powered force motor. The Active Combustor Control element developed intelligent staging and control strategies for low emission combustors. This included development and tests of a Controlled Pressure Fuel Nozzle for fuel sequencing, a Fuel Multiplexer for individual fuel cup metering, and model-based control logic. Both the Hydraulic Multiplexer and Controlled Pressure Fuel Nozzle system were cleared for engine test. The Fuel Multiplexer was cleared for combustor rig test which must be followed by an engine test to achieve full maturation.

Airborne rotary separator study

NASA Astrophysics Data System (ADS)

Drnevich, R. F.; Nowobilski, J. J.

1992-12-01

Several air breathing propulsion concepts for future earth-to-orbit transport vehicles utilize air collection and enrichment, and subsequent storage of liquid oxygen for later use in the vehicle mission. Work performed during the 1960's established the feasibility of substantially reducing weight and volume of a distillation type air separator system by operating the distillation elements in high 'g' fields obtained by rotating the separator assembly. The purpose of this study was to evaluate various fuels and fuel combinations with the objective of minimizing the weight and increase the ready alert capability of the plane. Fuels will be used to provide energy as well as act as heat sinks for the on-board heat rejection system. Fuel energy was used to provide power for air separation as well as to produce refrigeration for liquefaction of oxygen enriched air, besides its primary purpose of vehicle propulsion. The heat generated in the cycle was rejected to the fuel and water which is also carried on board the vehicle.The fuels that were evaluated include JP4, methane, and hydrogen. Hydrogen served as a comparison to the JP4 and methane cases.

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

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

D. L. Chichester; S. J. Thompson

2013-09-01

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium inmore » the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

DOE PAGES

David, Stan A.; Miller, Roger G.; Feng, Zhili

2016-08-31

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

Welding of unique and advanced alloys for space and high-temperature applications: welding and weldability of iridium and platinum alloys

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

David, Stan A.; Miller, Roger G.; Feng, Zhili

Advances have been made in developing alloys for space power systems for spacecraft that travel long distances to various planets. The spacecraft are powered by radioisotope thermoelectric generators (RTGs) and the fuel element in RTGs is plutonia. For safety and containment of the radioactive fuel element, the heat source is encapsulated in iridium or platinum alloys. Ir and Pt alloys are the alloys of choice for encapsulating radioisotope fuel pellets. Ir and Pt alloys were chosen because of their high-temperature properties and compatibility with the oxide fuel element and the graphite impact shells. This review addresses the alloy design andmore » welding and weldability of Ir and Pt alloys for use in RTGs.« less

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

DOEpatents

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

2010-02-23

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

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

DOEpatents

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

2011-03-01

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

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

DOEpatents

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

2013-09-03

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

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Progress in the Development of Oxygen Reduction Reaction Catalysts for Low-Temperature Fuel Cells

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

Li, Dongguo; Lv, Haifeng; Kang, Yijin

2016-04-06

In this paper, we present a brief summary on the most recent progress in the design of catalysts for electrochemical reduction of oxygen. The main challenge in the wide spread of fuel cell technology is to lower the content of, or even eliminate, Pt and other precious metals in catalysts without sacrificing their performance. Pt-based nanosized catalysts with novel and refined architectures continue to dominate in catalytic performance, and formation of Pt-skin-like surfaces is key to achieving the highest values in activity. Moreover, durability has also been improved in Pt-based systems with addition of Au, which plays an important rolemore » in stabilizing the Pt topmost layers against dissolution. However, various carbon-based materials without precious metal have shown improvement in activity and durability and have been explored to serve as catalyst supports. Finally, understanding how the doped elements interact with each other and/or carbon is challenging and necessary in the design of robust fuel cell catalysts.« less

Advances in proton-exchange membranes for fuel cells: an overview on proton conductive channels (PCCs).

PubMed

Wu, Liang; Zhang, Zhenghui; Ran, Jin; Zhou, Dan; Li, Chuanrun; Xu, Tongwen

2013-04-14

Proton-exchange membranes (PEM) display unique ion-selective transport that has enabled a breakthrough in high-performance proton-exchange membrane fuel cells (PEMFCs). Elemental understanding of the morphology and proton transport mechanisms of the commercially available Nafion® has promoted a majority of researchers to tune proton conductive channels (PCCs). Specifically, knowledge of the morphology-property relationship gained from statistical and segmented copolymer PEMs has highlighted the importance of the alignment of PCCs. Furthermore, increasing efforts in fabricating and aligning artificial PCCs in field-aligned copolymer PEMs, nanofiber composite PEMs and mesoporous PEMs have set new paradigms for improvement of membrane performances. This perspective profiles the recent development of the channels, from the self-assembled to the artificial, with a particular emphasis on their formation and alignment. It concludes with an outlook on benefits of highly aligned PCCs for fuel cell operation, and gives further direction to develop new PEMs from a practical point of view.

File-Based One-Way BISON Coupling Through VERA: User's Manual

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

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention [1–6]. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code [8,9] being developed by the Idaho National Laboratory (INL) is the focus of ongoing work in themore » Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework [10], BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, this work uses a 2D azimuthally symmetric (R-Z) smeared-pellet model. This manual is intended to cover (1) the procedure pertaining to the standalone BISON one-way coupling from VERA and (2) the procedure to generate BISON fuel temperature tables that VERA can use.« less

Biomass Burning: Energy and Emissions Performance of Traditional and Improved Cookstoves Under Controlled Laboratory Conditions

NASA Astrophysics Data System (ADS)

Arora, Pooja

Indoor air pollution (IAP) from solid biomass fuel burning in traditional cookstoves is a leading problem all the world which is responsible for health and climate related impacts. The immediate solution in order to combat this threat has been introduction of improved cookstoves among rural populations who doesn't have access clean energy. The extent of improvement in new cookstove designs, in terms of higher energy efficiency and lower emissions is in turn dependent on the customary behaviour of the users on field. The field based cookstove testing conducted in various studies show a disagreement between performance measures in the lab and real world conditions. Some of the important variables which reflect the actual user behaviour on field depending on geographical location include fuel characteristics and cooking cycle. In this thesis, the research approach focused on user-centred testing methodology for cookstoves. The variation in cookstove performance in terms of energy and emission parameters was assessed by isolating the impact of individual variables i.e. types of fuel and cooking cycles. The energy parameters which served as indicators of cookstove performance included SEC and power input, and EFs for CO and PM were used as emission parameters. PM emissions were further analysed with the help of physical and chemical characterization studies. The physical characterization focused on size distribution of the particulate using optical and electron microscopy techniques. While chemical characterization was conducted using quantification methods for organic and elemental carbon content of PM using TOR and CBMS techniques. The test variables were identified through field survey and literature review and were replicated under controlled laboratory conditions where emissions were sampled using hood method. The research resulted in six research papers addressing specific hypothesis related the problem identified through literature survey. The results showed that the difference in cooking cycles and fuel usage pattern has a significant impact on the overall performance of cookstoves. The study provided an in depth analysis of the difference in combustion conditions during the simulated and actual cookstove cooking conditions. The cooking cycle of two different regions of India indicated a significant change in cookstove performance in terms of CO and PM emissions in addition to energy consumption. It was also observed that among the different mix fuel conditions (combinations of wood, crop residue and cow dung), wood when used as a single fuel resulted in lowest CO and PM emissions. This was again a misrepresentation of field performance (specifically for regions with wood scarcity) where use of mix fuels is a prevalent practice followed while cooking. The impact of these variables was also significant on chemical characteristics of PM in terms of elemental carbon (EC)/black carbon (BC) and organic carbon (BC). Similar deviations were also found in physical characteristics of PM where PM size taken as test parameter. Therefore, relying on simulated test protocols might not represent the actual cookstove performance that might exist on field. This calls for an immediate attention towards the development of comprehensive test guidelines for cookstoves which not only highlights the technical aspects but also the social preferences of the targeted rural populations.

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

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

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

2014-04-23

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

Space storable propellant performance program coaxial injector characterization

NASA Technical Reports Server (NTRS)

Burick, R. J.

1972-01-01

An experimental program was conducted to characterize the circular coaxial injector concept for application with the space-storable gas/liquid propellant combination FLOX(82.6% F2)/CH4(g) at high pressure. The primary goal of the program was to obtain high characteristic velocity efficiency in conjunction with acceptable injector/chamber compatibility. A series of subscale (single element) cold flow and hot fire experiments was employed to establish design criteria for a 3000-lbf (sea level) engine operating at 500 psia. The subscale experiments characterized both high performance core elements and peripheral elements with enhanced injector/chamber compatibility. The full-scale injector which evolved from the study demonstrated a performance level of 99 percent of the theoretical shifting characteristic exhaust velocity with low chamber heat flux levels. A 44-second-duration firing demonstrated the durability of the injector. Parametric data are presented that are applicable for the design of circular, coaxial injectors that operate with injection dynamics (fuel and oxidizer velocity, etc.) similar to those employed in the work reported.

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, January 1, 1963--March 31, 1963

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT ...

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

34. DETAILS AND SECTIONS OF SHIELDING TANK FUEL ELEMENT SUPPORT FRAME. F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-4. INEL INDEX CODE NUMBER: 075 0701 60 851 151978. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

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)

JACKETED URANIUM SLUG

DOEpatents

Ohlinger, L.A.; Cooper, C.M.

1958-10-01

Fuel elements for nuclear reactors are described. Eacb fuel element is comprised of a solid cylindrical slug containing fissionable material enclosed within a fluid tight jacket of neutron permeable material such as aluminum. The jacket is provided with a flexible end cap and with a sealing member having a substantially fluid-tight fit within the jacket in tight abutment with the end cap and the end of the slug. A fluid passage is provided between the end of the slug and the cap whereby leakage fiuid is principally directed to the end of the slug. In this manner, any reaction between the fissionable material and fiuid which may take place occurs more rapidly at the end of the slug than along the sides between the slug and the jacket, thereby causing longitudinal expansion of the fuel element prior to radial expansion. The longitudinal expansion can be readily detected and the fuel element removed from the coolant tube before radial expansion causes it to become jammed in the tube.

Chemical Dissolution of Simulant FCA Cladding and Plates

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

Daniel, G.; Pierce, R.; O'Rourke, P.

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3-KF) flowsheets ofmore » H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.« 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.

Emission Performance of Low Cetane Naphtha as Drop-In Fuel on a Multi-Cylinder Heavy-Duty Diesel Engine and Aftertreatment System

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

LeePhD, John; TzanetakisPhD, Tom; Travers, Michael

With higher volatility and longer ignition delay characteristics than typical diesel fuel, low cetane naphtha fuel has been shown to promote partially premixed combustion and produce lower soot for improved fuel economy. In this study, emission performance of low cetane, low octane naphtha (CN 35, RON 60) as a drop-in fuel was examined on a MY13 Cummins ISX15 6-cylinder heavy-duty on-highway truck engine and aftertreatment system. Using the production hardware and development calibrations, both the engine-out and tailpipe emissions of naphtha and ultra-low sulfur diesel (ULSD) fuels were examined during the EPA s heavy-duty emission testing cycles. Without any modificationmore » to the calibrations, the tailpipe emissions were comparable when using naphtha or ULSD on the heavy duty Federal Test Procedure (FTP) and ramped modal cycle (RMC) test cycles. Overall lower CO2 emissions and fuel consumption were also measured for naphtha due in part to its higher heating value and higher hydrogen to carbon ratio. Engine-out and tailpipe NOx emissions were lower for naphtha fuel at the same catalyst conversion levels and measured particulate matter (PM) emissions were also lower when using naphtha due to its higher volatility and lower aromatic content compared to ULSD. To help assess the potential impact on diesel particulate filter design and operation, engine-out PM samples were collected and characterized at the B50 operating point. A significant reduction in elemental carbon (EC) within the particulate emissions was found when using naphtha compared to ULSD.« less

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

NASA Astrophysics Data System (ADS)

Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

2016-11-01

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

DOE PAGES

Williamson, R. L.; Capps, N. A.; Liu, W.; ...

2016-09-27

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) ormore » plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.« less

A Scalable, Parallel Approach for Multi-Point, High-Fidelity Aerostructural Optimization of Aircraft Configurations

NASA Astrophysics Data System (ADS)

Kenway, Gaetan K. W.

This thesis presents new tools and techniques developed to address the challenging problem of high-fidelity aerostructural optimization with respect to large numbers of design variables. A new mesh-movement scheme is developed that is both computationally efficient and sufficiently robust to accommodate large geometric design changes and aerostructural deformations. A fully coupled Newton-Krylov method is presented that accelerates the convergence of aerostructural systems and provides a 20% performance improvement over the traditional nonlinear block Gauss-Seidel approach and can handle more exible structures. A coupled adjoint method is used that efficiently computes derivatives for a gradient-based optimization algorithm. The implementation uses only machine accurate derivative techniques and is verified to yield fully consistent derivatives by comparing against the complex step method. The fully-coupled large-scale coupled adjoint solution method is shown to have 30% better performance than the segregated approach. The parallel scalability of the coupled adjoint technique is demonstrated on an Euler Computational Fluid Dynamics (CFD) model with more than 80 million state variables coupled to a detailed structural finite-element model of the wing with more than 1 million degrees of freedom. Multi-point high-fidelity aerostructural optimizations of a long-range wide-body, transonic transport aircraft configuration are performed using the developed techniques. The aerostructural analysis employs Euler CFD with a 2 million cell mesh and a structural finite element model with 300 000 DOF. Two design optimization problems are solved: one where takeoff gross weight is minimized, and another where fuel burn is minimized. Each optimization uses a multi-point formulation with 5 cruise conditions and 2 maneuver conditions. The optimization problems have 476 design variables are optimal results are obtained within 36 hours of wall time using 435 processors. The TOGW minimization results in a 4.2% reduction in TOGW with a 6.6% fuel burn reduction, while the fuel burn optimization resulted in a 11.2% fuel burn reduction with no change to the takeoff gross weight.

Fuel cell generator energy dissipator

DOEpatents

Veyo, Stephen Emery; Dederer, Jeffrey Todd; Gordon, John Thomas; Shockling, Larry Anthony

2000-01-01

An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a fuel cell generator when the electrical power output of the fuel cell generator is terminated. During a generator shut down condition, electrically resistive elements are automatically connected across the fuel cell generator terminals in order to draw current, thereby depleting the fuel

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

NASA Astrophysics Data System (ADS)

Mella, R.; Wenman, M. R.

2013-06-01

Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of different FE and finite difference models. Non-linear mechanical behaviour of the fuel and cladding including, fuel creep and swelling and cladding creep and plasticity each with dependencies on a variety of different properties. A fission gas release model which takes inputs from first principles calculations. Explicitly integrated inventory calculations performed in a coupled manner. Freedom to model steady state and transient behaviour using implicit time integration. The whole pin geometry is considered over an entire typical fuel life. The model showed by examination of normal operation and a subsequent transient chosen for software demonstration purposes: ABAQUS may be a sufficiently flexible platform to develop a complete and verified fuel performance code. The importance and effectiveness of the geometry of the fuel spacer pellets was characterised. The fuels performance under normal conditions (high friction no power spikes) would not suggest serious degradation of the cladding in fuel life. Large plastic strains were found when pellet bonding was strong, these would appear at all pellets cladding triple points and all pellet radial crack and cladding interfaces thus showing a possible axial direction to cracks forming from ductility exhaustion.

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

DOEpatents

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

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

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

SHAPED FISSIONABLE METAL BODIES

DOEpatents

Wigner, E.P.; Williamson, R.R.; Young, G.J.

1958-10-14

A technique is presented for grooving the surface of fissionable fuel elements so that expansion can take place without damage to the interior structure of the fuel element. The fissionable body tends to develop internal stressing when it is heated internally by the operation of the nuclear reactor and at the same time is subjected to surface cooling by the circulating coolant. By producing a grooved or waffle-like surface texture, the annular lines of tension stress are disrupted at equally spaced intervals by the grooves, thereby relieving the tension stresses in the outer portions of the body while also facilitating the removal of accumulated heat from the interior portion of the fuel element.

Parametric Design of Injectors for LDI-3 Combustors

NASA Technical Reports Server (NTRS)

Ajmani, Kumud; Mongia, Hukam; Lee, Phil

2015-01-01

Application of a partially calibrated National Combustion Code (NCC) for providing guidance in the design of the 3rd generation of the Lean-Direct Injection (LDI) multi-element combustion configuration (LDI-3) is summarized. NCC was used to perform non-reacting and two-phase reacting flow computations on several LDI-3 injector configurations in a single-element and a five-element injector array. All computations were performed with a consistent approach for mesh-generation, turbulence, spray simulations, ignition and chemical kinetics-modeling. Both qualitative and quantitative assessment of the computed flowfield characteristics of the several design options led to selection of an optimal injector LDI- 3 design that met all the requirements including effective area, aerodynamics and fuel-air mixing criteria. Computed LDI-3 emissions (namely, NOx, CO and UHC) will be compared with the prior generation LDI- 2 combustor experimental data at relevant engine cycle conditions.

Benchmarking criticality analysis of TRIGA fuel storage racks.

PubMed

Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

2017-01-01

A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

Particle Morphology and Elemental Composition of Smoke Generated by Overheating Common Spacecraft Materials

NASA Technical Reports Server (NTRS)

Meyer, Marit E.

2015-01-01

Fire safety in the indoor spacecraft environment is concerned with a unique set of fuels which are designed to not combust. Unlike terrestrial flaming fires, which often can consume an abundance of wood, paper and cloth, spacecraft fires are expected to be generated from overheating electronics consisting of flame resistant materials. Therefore, NASA prioritizes fire characterization research for these fuels undergoing oxidative pyrolysis in order to improve spacecraft fire detector design. A thermal precipitator designed and built for spacecraft fire safety test campaigns at the NASA White Sands Test Facility (WSTF) successfully collected an abundance of smoke particles from oxidative pyrolysis. A thorough microscopic characterization has been performed for ten types of smoke from common spacecraft materials or mixed materials heated at multiple temperatures using the following techniques: SEM, TEM, high resolution TEM, high resolution STEM and EDS. Resulting smoke particle morphologies and elemental compositions have been observed which are consistent with known thermal decomposition mechanisms in the literature and chemical make-up of the spacecraft fuels. Some conclusions about particle formation mechanisms are explored based on images of the microstructure of Teflon smoke particles and tar ball-like particles from Nomex fabric smoke.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

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

1963-07-01

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

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

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

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

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

2012-07-01

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

The EPA National Fuels Surveillance Network. I. Trace constituents in gasoline and commercial gasoline fuel additives.

PubMed Central

Jungers, R H; Lee, R E; von Lehmden, D J

1975-01-01

A National Fuels Surveillance Network has been established to collect gasoline and other fuels through the 10 regional offices of the Environmental Protection Agency. Physical, chemical, and trace element analytical determinations are made on the collected fuel samples to detect components which may present an air pollution hazard or poison exhaust catalytic control devices. A summary of trace elemental constituents in over 50 gasoline samples and 18 commercially marketed consumer purchased gasoline additives is presented. Quantities of Mn, Ni, Cr, Zn, Cu, Fe, Sb, B, Mg, Pb, and S were found in most regular and premium gasoline. Environmental implications of trace constituents in gasoline are discussed. PMID:1157783

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

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

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Improved coking resistance of direct ethanol solid oxide fuel cells with a Ni-Sx anode

NASA Astrophysics Data System (ADS)

Yan, Ning; Luo, Jing-Li; Chuang, Karl T.

2014-03-01

In this study, the coking resistance of anode supported direct ethanol solid oxide fuel cell with a Ni-Sx anode was investigated comparatively with the conventional cell using pure Ni catalyst. The surface catalytic properties of Ni were manipulated via depositing a layer of S atoms. It was confirmed that on the surface of Ni, a combination of S monolayer and elemental S was formed without producing Ni3S2 phase. The developed Ni-Sx cell exhibited a significantly improved coke resistivity in ethanol feed while maintaining an adequately high performance. The S species on Ni enabled the suppression of the coke formation as well as the alleviation of the metal dusting effect of the anode structure. After operating in ethanol fuel for identical period of time at 850 °C, a maximum power density of 400 mW cm-2 was sustained whereas the conventional cell performance decreased to less than 40 mW cm-2 from the original 704 mW cm-2. In an optimized stability test, the Ni-Sx cell operated at 750 °C for more than 22 h until the fuel drained without any degradation.

Modeling the Pore Formation Mechanism in UMo/AL Dispersion Fuel

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

Kim, Yeon Soo; Jamison, L.; Hofman, G.

In UMo/Al dispersion fuel meat, pores formed in the ILs or at IL-Al interfaces tend to increase in size with irradiation, potentially limiting performance of this fuel. There has been no universally accepted mechanism for the formation and growth of this type of pore. However, there is a consensus that the stress state determined by meat swelling and fission- induced creep is one of the determinants, and fission gas availability at the pore site is another. Five dispersion RERTR miniplates that have well defined irradiation conditions and PIE data were selected for examination. Meat swelling and pore volume were measuredmore » in each plate. ABAQUS finite element analysis (FEA) package was utilized to obtain the time-dependent evolution of mechanical states in the plates while matching the measured meat swelling and creep. Interpretation of these results give insights on how to model a failure function – a predictor for large pore formation – using variables such as meat swelling, interaction layer growth, stress, and creep. This model can be used for optimizing fuel design parameters to reach the desired goal: meeting high power and performance reactor demand.« less

Axially staggered seed-blanket reactor fuel module construction

DOEpatents

Cowell, Gary K.; DiGuiseppe, Carl P.

1985-01-01

A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-07-11

Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

Autothermal reforming catalyst having perovskite structure

DOEpatents

Krumpel, Michael [Naperville, IL; Liu, Di-Jia [Naperville, IL

2009-03-24

The invention addressed two critical issues in fuel processing for fuel cell application, i.e. catalyst cost and operating stability. The existing state-of-the-art fuel reforming catalyst uses Rh and platinum supported over refractory oxide which add significant cost to the fuel cell system. Supported metals agglomerate under elevated temperature during reforming and decrease the catalyst activity. The catalyst is a perovskite oxide or a Ruddlesden-Popper type oxide containing rare-earth elements, catalytically active firs row transition metal elements, and stabilizing elements, such that the catalyst is a single phase in high temperature oxidizing conditions and maintains a primarily perovskite or Ruddlesden-Popper structure under high temperature reducing conditions. The catalyst can also contain alkaline earth dopants, which enhance the catalytic activity of the catalyst, but do not compromise the stability of the perovskite structure.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

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

Navarro, Jorge

2013-12-01

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent tomore » the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.« less

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

Validating the BISON fuel performance code to integral LWR experiments

DOE PAGES

Williamson, R. L.; Gamble, K. A.; Perez, D. M.; ...

2016-03-24

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to datemore » for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.« less

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

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

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

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

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

Production test IP-544-A, irradiation of 1.6% enriched thick walled single tube elements in KER-1 and 2

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

Kratzer, W.K.; Wise, M.J.

1962-12-12

The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

Digital image processing techniques for the analysis of fuel sprays global pattern

NASA Astrophysics Data System (ADS)

Zakaria, Rami; Bryanston-Cross, Peter; Timmerman, Brenda

2017-12-01

We studied the fuel atomization process of two fuel injectors to be fitted in a new small rotary engine design. The aim was to improve the efficiency of the engine by optimizing the fuel injection system. Fuel sprays were visualised by an optical diagnostic system. Images of fuel sprays were produced under various testing conditions, by changing the line pressure, nozzle size, injection frequency, etc. The atomisers were a high-frequency microfluidic dispensing system and a standard low flow-rate fuel injector. A series of image processing procedures were developed in order to acquire information from the laser-scattering images. This paper presents the macroscopic characterisation of Jet fuel (JP8) sprays. We observed the droplet density distribution, tip velocity, and spray-cone angle against line-pressure and nozzle-size. The analysis was performed for low line-pressure (up to 10 bar) and short injection period (1-2 ms). Local velocity components were measured by applying particle image velocimetry (PIV) on double-exposure images. The discharge velocity was lower in the micro dispensing nozzle sprays and the tip penetration slowed down at higher rates compared to the gasoline injector. The PIV test confirmed that the gasoline injector produced sprays with higher velocity elements at the centre and the tip regions.

Fuel Economy Label and CAFE Data Inventory

EPA Pesticide Factsheets

The Fuel Economy Label and CAFE Data asset contains measured summary fuel economy estimates and test data for light-duty vehicle manufacturers by model for certification as required under the Energy Policy and Conservation Act of 1975 (EPCA) and The Energy Independent Security Act of 2007 (EISA) to collect vehicle fuel economy estimates for the creation of Economy Labels and for the calculation of Corporate Average Fuel Economy (CAFE). Manufacturers submit data on an annual basis, or as needed to document vehicle model changes.The EPA performs targeted fuel economy confirmatory tests on approximately 15% of vehicles submitted for validation. Confirmatory data on vehicles is associated with its corresponding submission data to verify the accuracy of manufacturer submissions beyond standard business rules. Submitted data comes in XML format or as documents, with the majority of submissions being sent in XML, and includes descriptive information on the vehicle itself, fuel economy information, and the manufacturer's testing approach. This data may contain proprietary information (CBI) such as information on estimated sales or other data elements indicated by the submitter as confidential. CBI data is not publically available; however, within the EPA data can accessed under the restrictions of the Office of Transportation and Air Quality (OTAQ) CBI policy [RCS Link]. Datasets are segmented by vehicle model/manufacturer and/or year with corresponding fuel economy, te

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Astrophysics Data System (ADS)

Emrich, William J.

2008-01-01

To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

Chemical characterization of the fine particle emissions from commercial aircraft engines during the Aircraft Particle Emissions eXperiment (APEX) 1 to 3.

PubMed

Kinsey, J S; Hays, M D; Dong, Y; Williams, D C; Logan, R

2011-04-15

This paper addresses the need for detailed chemical information on the fine particulate matter (PM) generated by commercial aviation engines. The exhaust plumes of seven turbofan engine models were sampled as part of the three test campaigns of the Aircraft Particle Emissions eXperiment (APEX). In these experiments, continuous measurements of black carbon (BC) and particle surface-bound polycyclic aromatic compounds (PAHs) were conducted. In addition, time-integrated sampling was performed for bulk elemental composition, water-soluble ions, organic and elemental carbon (OC and EC), and trace semivolatile organic compounds (SVOCs). The continuous BC and PAH monitoring showed a characteristic U-shaped curve of the emission index (EI or mass of pollutant/mass of fuel burned) vs fuel flow for the turbofan engines tested. The time-integrated EIs for both elemental composition and water-soluble ions were heavily dominated by sulfur and SO(4)(2-), respectively, with a ∼2.4% median conversion of fuel S(IV) to particle S(VI). The corrected OC and EC emission indices obtained in this study ranged from 37 to 83 mg/kg and 21 to 275 mg/kg, respectively, with the EC/OC ratio ranging from ∼0.3 to 7 depending on engine type and test conditions. Finally, the particle SVOC EIs varied by as much as 2 orders of magnitude with distinct variations in chemical composition observed for different engine types and operating conditions.

Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2008-01-01

To support the eventual development of a nuclear thermal rocket engine, a state-of-the-art experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

Nonlinear heat transfer and structural analyses of SSME turbine blades

NASA Technical Reports Server (NTRS)

Abdul-Aziz, A.; Kaufman, A.

1987-01-01

Three-dimensional nonlinear finite-element heat transfer and structural analyses were performed for the first stage high-pressure fuel turbopump blade of the space shuttle main engine (SSME). Directionally solidified (DS) MAR-M 246 material properties were considered for the analyses. Analytical conditions were based on a typical test stand engine cycle. Blade temperature and stress-strain histories were calculated using MARC finite-element computer code. The study was undertaken to assess the structural response of an SSME turbine blade and to gain greater understanding of blade damage mechanisms, convective cooling effects, and the thermal-mechanical effects.

Tribology and energy efficiency: from molecules to lubricated contacts to complete machines.

PubMed

Taylor, Robert Ian

2012-01-01

The impact of lubricants on energy efficiency is considered. Molecular details of base oils used in lubricants can have a great impact on the lubricant's physical properties which will affect the energy efficiency performance of a lubricant. In addition, molecular details of lubricant additives can result in significant differences in measured friction coefficients for machine elements operating in the mixed/boundary lubrication regime. In single machine elements, these differences will result in lower friction losses, and for complete systems (such as cars, trucks, hydraulic circuits, industrial gearboxes etc.) lower fuel consumption or lower electricity consumption can result.

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

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

Unal, Cetin; Galloway, Jack D.

2014-09-12

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

TWISTED RIBBON FUEL ELEMENT

DOEpatents

Breden, C.R.; Schultz, A.B.

1961-06-01

A reactor core formed of bundles of parallel fuel elements in the form of ribbons is patented. The fuel ribbons are twisted about their axes so as to have contact with one another at regions spaced lengthwise of the ribbons and to be out of contact with one another at locations between these spaced regions. The contact between the ribbons is sufficient to allow them to be held together in a stable bundle in a containing tube without intermediate support, while permitting enough space between the ribbon for coolant flowing.

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.

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)

Chemical and sewage sludge co-incineration in a full-scale MSW incinerator: toxic trace element mass balance.

PubMed

Biganzoli, Laura; Grosso, Mario; Giugliano, Michele; Campolunghi, Manuel

2012-10-01

Co-incineration of sludges with MSW is a quite common practice in Europe. This paper illustrates a case of co-incineration of both sewage sludges and chemical sludges, the latter obtained from drinking water production, in a waste-to-energy (WTE) plant located in northern Italy and equipped with a grate furnace, and compares the toxic trace elements mass balance with and without the co-incineration of sludges. The results show that co-incineration of sewage and chemical sludges does not result in an increase of toxic trace elements the total release in environment, with the exception of arsenic, whose total release increases from 1 mg t(fuel) (-1) during standard operation to 3 mg t(fuel) (-1) when sludges are co-incinerated. The increase of arsenic release is, however, attributable to the sole bottom ashes, where its concentration is five times higher during sludge co-incineration. No variation is observed for arsenic release at the stack. This fact is a further guarantee that the co-incineration of sludges, when performed in a state-of-the-art WTE plant, does not have negative effects on the atmospheric environment.

A Very Large Eddy Simulation of the Nonreacting Flow in a Single-Element Lean Direct Injection Combustor Using PRNS with a Nonlinear Subscale Model

NASA Technical Reports Server (NTRS)

Shih, Tsan-Hsing; Liu, Nan-Suey

2009-01-01

Very large eddy simulation (VLES) of the nonreacting turbulent flow in a single-element lean direct injection (LDI) combustor has been successfully performed via the approach known as the partially resolved numerical simulation (PRNS/VLES) using a nonlinear subscale model. The grid is the same as the one used in a previous RANS simulation, which was considered as too coarse for a traditional LES simulation. In this study, we first carry out a steady RANS simulation to provide the initial flow field for the subsequent PRNS/VLES simulation. We have also carried out an unsteady RANS (URANS) simulation for the purpose of comparing its results with that of the PRNS/VLES simulation. In addition, these calculated results are compared with the experimental data. The present effort has demonstrated that the PRNS/VLES approach, while using a RANS type of grid, is able to reveal the dynamically important, unsteady large-scale turbulent structures occurring in the flow field of a single-element LDI combustor. The interactions of these coherent structures play a critical role in the dispersion of the fuel, hence, the mixing between the fuel and the oxidizer in a combustor.

Chemical characterization of biomass fuel smoke particles of rural kitchens of South Asia

NASA Astrophysics Data System (ADS)

Deka, Pratibha; Hoque, Raza Rafiqul

2015-05-01

Biomass fuel smoke particles (BFSPs) of rural kitchens collected during dry and wet seasons were characterized for elements, anions and carbon. The BFSPs of kitchens using varied biomass fuel types viz. cow dung stick, mixed biomass, cow-dung stick-mixed biomass and sugarcane bagasse were chosen for the study. The BFSPs from cow dung fuel stick showed higher levels of elements, anions and particulate carbon than other BFSPs. Calcium, K, Fe and Mg were the major elements found in all BFSPs, which did not vary much between the seasons. Sulphate was found to be the dominant anion present in all BFSPs followed by Clˉ and PO43-. Seasonal variation was pronounced in the case of abundance of anions and particulate carbon. The ratio OC/EC, often used as source signature of biomass burning, was found to be within 1.89-7.41 and 1.72-6.19 during dry and wet seasons respectively.

ON CRITICAL MASS ANALYSIS OF JRR-2

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

None

1961-01-01

The critica mass of the JRR-2 was found to be 15 fuel elements, instead of 8 as expected, when the reactor reached criticaity. The critica mass was analyzed by AMF and JAERI a few years ago, but afterwards some modifications have been made of the stucture for the reinforcement, for example, during the construction. The critical mass is recalculated perfectly and the difference bctween 15 and S fuel elements is discussed. The deviation of the critical mass is mainly caused by the effects of control rods, fuel elcments, grid-plate, etc., in the reflector; only heavy water or light water wasmore » conaidered as the reflector in the previous calculation. A simple method is used to calculate the critical mass. The effective multiplication factor for the core with 15 fuel elements is obtained about 2% higher than the experimental value. This difference is also discussed in detail. (auth)« less

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT ...

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

36. DETAILS AND SECTIONS OF SHIELDING TANK, FUEL ELEMENT SUPPORT FRAME AND SUPPORT PLATFORM, AND SAFETY MECHANISM ASSEMBLY (SPRING-LOADED HINGE). F.C. TORKELSON DRAWING NUMBER 842-ARVFS-701-S-1. INEL INDEX CODE NUMBER: 075 0701 60 851 151975. - Idaho National Engineering Laboratory, Advanced Reentry Vehicle Fusing System, Scoville, Butte County, ID

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

Performance of nickel-based oxygen carrier produced using renewable fuel aloe vera

NASA Astrophysics Data System (ADS)

Afandi, NF; Devaraj, D.; Manap, A.; Ibrahim, N.

2017-04-01

Consuming and burning of fuel mainly fossil fuel has gradually increased in this upcoming era due to high-energy demand and causes the global warming. One of the most effective ways to reduce the greenhouse gases is by capturing carbon dioxide (CO2) during the combustion process. Chemical looping combustion (CLC) is one of the most effective methods to capture the CO2 without the need of an energy intensive air separation unit. This method uses oxygen carrier to provide O2 that can react with fuel to form CO2 and H2O. This research focuses on synthesizing NiO/NiAl2O4 as an oxygen carrier due to its properties that can withstand high temperature during CLC application. The NiO/NiAl2O4 powder was synthesized using solution combustion method with plant extract renewable fuel, aloe vera as the fuel. In order to optimize the performance of the particles that can be used in CLC application, various calcination temperatures were varied at 600°C, 800°C, 1050°C and 1300°C. The phase and morphology of obtained powders were characterized using X-ray diffraction (XRD) and Field Emission Microscopy (FESEM) respectively together with the powder elements. In CLC application, high reactivity can be achieved by using smaller particle size of oxygen carrier. This research succeeded in producing nano-structured powder with high crystalline structure at temperature 1050°C which is suitable to be used in CLC application.

Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

NASA Astrophysics Data System (ADS)

Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

1993-09-01

The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

Fuel assembly shaker and truck test simulation

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

Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.

2014-09-30

This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revisedmore » model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when traveling down the same road at the same speed. It is recommended that the SNL conveyance system used in testing be characterized through modal analysis and frequency response analysis to provide context and assist in the interpretation of the strain data that was collected during the truck test campaign.« less

GEH-4-42, 47; Hot pressed, I and E cooled fuel element irradiation

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

Neidner, R.

1959-11-02

In our continual effort to improve the present fuel elements which are irradiated in the numerous Hanford reactors, we have made what we believe to be a significant improvement in the hot pressing process for jacketing uranium fuel slugs. We are proposing a large scale evaluation testing program in the Hanford reactors but need the vital and basic information on the operating characteristics of this type slug under known and controlled operating conditions. We, therefore, have prepared two typical fuel slugs and will want them irradiated to about 1000 MWD/T exposure (this will require about four to five total cycles).

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.

An out-of-core thermionic-converter system for nuclear space power

NASA Technical Reports Server (NTRS)

Breitwieser, R.

1972-01-01

Design of the nuclear thermionic space power system, 40 50 70 Kw(e) power range, are given. The design configuration (1) meets the constraints of readily available launch vehicles; (2) allows for off-design operation including startup, shutdown, and possible emergency conditions; (3) provides tolerance of failure by extensive use of modular, redundant elements; (4) incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and (5) uses thermionic converters, nuclear fuel elements, and heat transfer devices in a geometrical form adapted from existing incore thermionic system designs. Designs and in some cases performance data for elements and groups of the elements of the system are included. Benefits of the highly modular system approach to reliability, safety, economy of development, and flexibility are discussed.

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

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

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

2016-01-30

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

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

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

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

Checking the possibility of controlling fuel element by X-ray computerized tomography

NASA Astrophysics Data System (ADS)

Trinh, V. B.; Zhong, Y.; Osipov, S. P.; Batranin, A. V.

2017-08-01

The article considers the possibility of checking fuel elements by X-ray computerized tomography. The checking tasks are based on the detection of particles of active material, evaluation of the heterogeneity of the distribution of uranium salts and the detection of clusters of uranium particles. First of all, scheme of scanning improve the performance and quality of the resulting three-dimensional images of the internal structure is determined. Further, the possibility of detecting clusters of uranium particles having the size of 1 mm3 and measuring the coordinates of clusters of uranium particles in the middle layer with the accuracy of within a voxel size (for the considered experiments of about 80 μm) is experimentally proved in the main part. The problem of estimating the heterogeneity of the distribution of the active material in the middle layer and the detection of particles of active material with a nominal diameter of 0.1 mm in the “blank” is solved.

NASA's Nuclear Thermal Propulsion Project

NASA Technical Reports Server (NTRS)

Houts, Michael; Mitchell, Sonny; Kim, Tony; Borowski, Stanley; Power, Kevin; Scott, John; Belvin, Anthony; Clement, Steven

2015-01-01

Space fission power systems can provide a power rich environment anywhere in the solar system, independent of available sunlight. Space fission propulsion offers the potential for enabling rapid, affordable access to any point in the solar system. One type of space fission propulsion is Nuclear Thermal Propulsion (NTP). NTP systems operate by using a fission reactor to heat hydrogen to very high temperature (>2500 K) and expanding the hot hydrogen through a supersonic nozzle. First generation NTP systems are designed to have an Isp of approximately 900 s. The high Isp of NTP enables rapid crew transfer to destinations such as Mars, and can also help reduce mission cost, improve logistics (fewer launches), and provide other benefits. However, for NTP systems to be utilized they must be affordable and viable to develop. NASA's Advanced Exploration Systems (AES) NTP project is a technology development project that will help assess the affordability and viability of NTP. Early work has included fabrication of representative graphite composite fuel element segments, coating of representative graphite composite fuel element segments, fabrication of representative cermet fuel element segments, and testing of fuel element segments in the Compact Fuel Element Environmental Tester (CFEET). Near-term activities will include testing approximately 16" fuel element segments in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES), and ongoing research into improving fuel microstructure and coatings. In addition to recapturing fuels technology, affordable development, qualification, and utilization strategies must be devised. Options such as using low-enriched uranium (LEU) instead of highly-enriched uranium (HEU) are being assessed, although that option requires development of a key technology before it can be applied to NTP in the thrust range of interest. Ground test facilities will be required, especially if NTP is to be used in conjunction with high value or crewed missions. There are potential options for either modifying existing facilities or constructing new ground test facilities. At least three potential options exist for reducing (or eliminating) the release of radioactivity into the environment during ground testing. These include fully containing the NTP exhaust during the ground test, scrubbing the exhaust, or utilizing an existing borehole at the Nevada National Security Site (NNSS) to filter the exhaust. Finally, the project is considering the potential for an early flight demonstration of an engine very similar to one that could be used to support human Mars or other ambitious missions. The flight demonstration could be an important step towards the eventual utilization of NTP.

An experimental investigation of hybrid kerosene burner configurations for TPV applications

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

Schroeder, K.L.; Rose, M.F.; Burkhalter, J.E.

1995-01-05

A key element in thermophotovoltaic power generation is the development of a compact and efficient configuration for the thermal source and emitter. In the present work, a hybrid configuration was investigated which was composed of a liquid fueled diffusion type burner utilizing the emitting or mantle structure as the combustion chamber. The prototype burner operates on kerosene at fuel flow rates up to 1.0 kg/hr. Fuel is atomized using an 78 kHz ultrasonic nozzle with multifuel capabilities. Combustion is stabilized and heat transfer is enhanced via forced recirculation interior to the mantle structures. These structures range in size from 600more » to 1200 cm{sup 3} and are porous in nature. This paper presents an introduction to issues specific to the use of small scale liquid fueled burners for TPV applications, and burner performance data for a series of configurations, in terms of combustor surface temperature distribution, maximum mass loading and efficiency. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}.« less

Advanced Fusion Reactors for Space Propulsion and Power Systems

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

Chapman, John J.

In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Protonmore » triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.« less

Advanced Fusion Reactors for Space Propulsion and Power Systems

NASA Technical Reports Server (NTRS)

Chapman, John J.

2011-01-01

In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

Results from a Field Trial of the Radio Frequency Based Cylinder Accountability and Tracking System at the Global Nuclear Fuel Americas Fuel Fabrication Facility

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

Fitzgerald, Peter; Laughter, Mark D; Martyn, Rose

The Cylinder Accountability and Tracking System (CATS) is a tool designed for use by the International Atomic Energy Agency (IAEA) to improve overall inspector efficiency through real-time unattended monitoring of cylinder movements, site specific rules-based event detection, and the capability to integrate many types of monitoring technologies. The system is based on the tracking of cylinder movements using (radio frequency) RF tags, and the collection of data, such as accountability weights, that can be associated with the cylinders. This presentation will cover the installation and evaluation of the CATS at the Global Nuclear Fuels (GNF) fuel fabrication facility in Wilmington,more » NC. This system was installed to evaluate its safeguards applicability, operational durability under operating conditions, and overall performance. An overview of the system design and elements specific to the GNF deployment will be presented along with lessons learned from the installation process and results from the field trial.« less

Effect of High Si Content on U3Si2 Fuel Microstructure

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Performance evaluation of a liquid tin anode solid oxide fuel cell operating under hydrogen, argon and coal

NASA Astrophysics Data System (ADS)

Khurana, Sanchit; LaBarbera, Mark; Fedkin, Mark V.; Lvov, Serguei N.; Abernathy, Harry; Gerdes, Kirk

2015-01-01

A liquid tin anode solid oxide fuel cell is constructed and investigated under different operating conditions. Electrochemical Impedance Spectroscopy (EIS) is used to reflect the effect of fuel feed as the EIS spectra changes significantly on switching the fuel from argon to hydrogen. A cathode symmetric cell is used to separate the impedance from the two electrodes, and the results indicate that a major contribution to the charge-transfer and mass-transfer impedance arises from the anode. The OCP of 0.841 V for the cell operating under argon as a metal-air battery indicates the formation of a SnO2 layer at the electrolyte/anode interface. The increase in the OCP to 1.1 V for the hydrogen fueled cell shows that H2 reduces the SnO2 film effectively. The effective diffusion coefficients are calculated using the Warburg element in the equivalent circuit model for the experimental EIS data, and the values of 1.9 10-3 cm2 s-1 at 700 °C, 2.3 10-3 cm2 s-1 at 800 °C and 3.5 10-3 cm2 s-1 at 900 °C indicate the system was influenced by diffusion of hydrogen in the system. Further, the performance degradation over time is attributed to the irreversible conversion of Sn to SnO2 resulting from galvanic polarization.

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

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

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

40 CFR 80.1141 - Small refinery exemption.

Code of Federal Regulations, 2010 CFR

2010-07-01

... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Renewable Fuel Standard § 80.1141 Small refinery exemption...)), is exempt from the renewable fuel standards of § 80.1105 and the requirements that apply to obligated... refinery application. The application must contain all of the elements required for small refinery...

Electric cartridge-type heater for producing a given non-uniform axial power distribution

DOEpatents

Clark, D.L.; Kress, T.S.

1975-10-14

An electric cartridge heater is provided to simulate a reactor fuel element for use in safety and thermal-hydraulic tests of model nuclear reactor systems. The electric heat-generating element of the cartridge heater consists of a specifically shaped strip of metal cut with variable width from a flat sheet of the element material. When spirally wrapped around a mandrel, the strip produces a coiled element of the desired length and diameter. The coiled element is particularly characterized by an electrical resistance that varies along its length due to variations in strip width. Thus, the cartridge heater is constructed such that it will produce a more realistic simulation of the actual nonuniform (approximately ''chopped'' cosine) power distribution of a reactor fuel element.