Development Status of CVD System to Deposit Tungsten onto UO2 Powder via the WCI6 Process
No abstract available
Mireles, O.↗
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Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion
No abstract available
Hickman, Robert↗
Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination
A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.
Collins, J. F.↗
Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C
The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.
Bowles, K. J.↗
Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication
Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.
Mireles, O. R.↗
A Historical Review of Cermet Fuel Development and the Engine Performance Implications
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.
Ceramic Nuclear Fuel↗
Design of a Uranium Dioxide Spheroidization System
The plasma spheroidization system (PSS) is the first process in the development of tungsten-uranium dioxide (W-UO2) fuel cermets. The PSS process improves particle spherocity and surface morphology for coating by chemical vapor deposition (CVD) process. Angular fully dense particles melt in an argon-hydrogen plasma jet at between 32-36 kW, and become spherical due to surface tension. Surrogate CeO2 powder was used in place of UO2 for system and process parameter development. Particles range in size from 100 - 50 microns in diameter. Student s t-test and hypothesis testing of two proportions statistical methods were applied to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders show great than 800% increase in the number of spherical particles over the stock powder with the mean spherocity only mildly improved. It is recommended that powders be processed two-three times in order to reach the desired spherocity, and that process parameters be optimized for a more narrow particles size range. Keywords: spherocity, spheroidization, plasma, uranium-dioxide, cermet, nuclear, propulsion
Cavender, Daniel P.↗
A Historical Review of Cermet Fuel Development and the Engine Performance Implications
To better understand Cermet engine performance, examined historical material development reports two issues: High vaporization rate of UO2, High temperature chemical stability of UO2. Cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance. Few samples were tested above 2770 K. Results above 2770 K are ambiguous. Contemporary testing may clarify performance. Cermet sample testing during the NERVA Rover era. Important properties, melting temperature, vaporization rate, strength, Brittle-to-Ductile Transition, cermet sample test results, engine performance, location, peak temperature.
Nuclear Propulsion↗
Results of a preliminary experimental investigation of a vapor transport fuel pin
In-Pile experimental tests were conducted on two fuel pins mounted side-by-side in a holder assembly. The fuel pins consisted of stacked fully enriched UO2 pellets enclosed in a type 316 stainless steel clad 1/2 inch outside diameter and 3 inches in length. Each fuel pin contained four pellets for a total of 22 grams of fuel weight. The experiment was operated to maintain the maximum clad temperature at 395 K corresponding to a fuel temperature at the central void surface of 2350 K. Total primary coolant water flow past the pin holder was calculated to be 110 gallons per minute. Forty-four gallons per minute at a velocity of 36.5 feet per second flowed around each fuel pin. Calculations indicated the maximum heat flux was at the point of the highest temperature and was equal to 0.24 kW/sq cm or 7.62 Btu/hr sq ft. A series of high resolution neutron radiographs of the nuclear fueled vapor transport capsule was taken. The thermal neutrons emitted from the core of the 60 megawatt Plum Brook reactor facility were used.
Lanzo, C. D.↗
In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode
The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.
Diianni, D. C.↗
Uranium-bearing minerals of lunar rock 12013.
The U distribution in rock 12013 was studied by fission track and elemental mapping techniques. Major U-bearing phases are whitlockite, apatite, zircon, and phase beta, which is a Zr-Ti mineral rich in Fe, Nb, Y, REE, and containing up to 3.6% UO2, 4.7% ThO2 and 4.2% PbO. Calculated microprobe ages for phase beta average 4.0 b.y. and are in reasonable agreement with Rb-Sr and K-Ar ages. Phase beta plays a significant role in the U-Th-Pb systematics of rock 12012 and may play a similar role in the model ages of lunar soil.
Haines, E. L.↗
Generation of a uranium plasma at near gaseous core reactor conditions.
A constricted sliding spark discharge is used to generate a high density, high temperature uranium plasma. Uranium particle densities up to 10 to the 20th power per cu cm are obtained over a temperature range of 30,000 to 50,000 K. The device consists of a capillary discharge channel lined with pressed and sintered UO2. A 250 joule capacitor bank is discharged into the channel, producing a plasma of 10-20 microsec duration. Spectroscopic observations are made over the spectral range of 1300 to 2500 A.
Davis, J. F., III↗
Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad
The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.
Saltsman, J. F.↗
Irradiation of TZM: Uranium dioxide fuel pin at 1700 K
A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.
Mcdonald, G. E.↗
Uranium dioxide fuel cladding strain investigation with the use of CYGRO-2 computer program
Previously irradiated UO2 thermionic fuel pins in which gross fuel-cladding strain occurred were modeled with the use of a computer program to define controlling parameters which may contribute to cladding strain. The computed strain was compared with measured strain, and the computer input data were studied in an attempt to get agreement with measured strain. Because of the limitations of the program and uncertainties in input data, good agreement with measured cladding strain was not attained. A discussion of these limitations is presented.
Smith, J. R.↗
Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K
Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.
Diianni, D. C.↗
Luminescence of powdered uranium glasses
Measurement of cathodoluminescence and photoluminescence efficiencies in powdered borosilicate glasses having different particle size and different uranium content. Excitation with 100 to 350 keV electrons and with 253.7 nm light was found to produce identical absolute radiant exitance spectra in powdered samples. The most efficient glass was one containing 29.4 wt% B2O3, 58.8 wt% SiO2, 9.8 wt% Na2O and 2.0 wt% UO2.
Eubanks, A. G.↗
Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins
Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.
Thoms, K. R.↗