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Radioisotope and Radiation Applications. Quarterly Progress Report

Description: An evaluation was given of the possible hazards to consumers from radioisotope residues in consumer products. A laboratory demonstration was given of the use of Mn/sup 54/ to facilitate removal of manganese from process feed water. lt was found in the hazards evaluation that the "worst case" of radiation exposure from residual radioisotopes in steel gives a radiation exposure somewhat less than the maximum allowable dose levels for occupational exposure. Initial study indicates that for actual cases, the radiation exposures to be expected from radioisotope residues in steel products would ordinarily be small compared to natural background. An exception to this generalization might be found when a longer lived isotope like Mn/sup 54/ was present. Preliminary results of the laboratory demonstnation of using Mn/sup 54/ to monitor the removal of manganese from feed water indicated that the method may allow a considerable improvement in accuracy of process control. The study of the mechanism of formation of free radicals in polymeric materials was continued. Emphasis was placed on examination of the effect of structural factors on the efficiency of free-radical site formation in acrylate polymers. The investigation was extended to include an examination of the effect on free-radical formation of the constituents on the carbon atom located alpha to the ester group. Polymethylacrylate, polymethylmethacrylate, and polymethyl- alpha -chloroacrylate were used in this study. Measurement of the volatile products from the irradiation of the polymethyl- alpha -chloroacrylate was completed. The data substantiated earlier findings which indicated that the point of attack in free-radical formation occurs on the ester side chain. (auth)
Date: January 18, 1961
Creator: Sunderman, D. N.
Partner: UNT Libraries Government Documents Department

Radioisotope and Radiation Applications. Quarterly Progress Report

Description: A series of experiments was completed using the process model for studying radiotracer control of iron removal from a nickel-refinery stream. The results of 13 experimental runs indicated that the radiotracer control concept is technically sound. The iron reduction ratios obtained by radioassay agreed well with the iron reduction ratios given by chemical analysis. The radioactivity monitor on the filtrate stream provided a rapid and sensitive indication of changes in process operating conditions. The survey of chemical engineering unit operations as potential intrinsic radiotracer applications was completed. The study of the mechanism of formation of free radicals in polymethacrylates was continued. Particular emphasis was placed on an examination of the effect of structural factors on the efficiency of free-radical site formation. The investigation of the influence of free-radical formation of the hydrocarbon constituent of the ester side chain was continued. In addition, polymer molecular weight was found to influence site concentration. (auth)
Date: October 31, 1960
Creator: Sunderman, D. N.
Partner: UNT Libraries Government Documents Department

High-Temperature Irradiation of Metals and Graphite in Flowing Helium

Description: As part of the coolant-core evaluation studies in an early phase of the Maritime Gas-Cooled Reactor program, nickel. K-Monel, niobium-1 wt.% zirconium, and graphite were irradiated in a convective flow of helium. A capsule of novel design was used to study the corrosion of the metals at 1500 and 960 ction prod- F in a 243-hr BRR experiment. Four heliumfilled quartz tori, containing the specimens, were encapsulated in a stainless steel capsule. A convective flow of helium was maintained in the closed tori by heating one leg of each torus to 1500 and the other at 960-F. Zirconium foilwas placed in two of the tori to getter gaseous impurities. No attack of the nickel and niobium-1 wt.% zirconium was observed at either temperature. Metallographic examination did show a 1 to 2.5- mil attack of the K-Monel at 1500 ction prod- F in a torus containing no getter. (auth)
Date: February 23, 1960
Creator: Miller, N. E.; Hamman, D. J.; Diethorn, W. S. & Goldthwaite, W. H.
Partner: UNT Libraries Government Documents Department

Effects of Irradiation on the Mechanical Properties of Tantalum

Description: Tensile, bend ductility, and hardness tests were performed at room temperature on irradiated tantalum sheet to determine the effect of irradiation on the strength and ductility. Sheet tensile specimens were irradiated in an attempt to produce corversions of tantalum to tungsten of approximately 1.5 and 3.0 wt.%. Unirradiated tantalum and arc-melted alloys of tantalum-1.5 and -3 wt.% tungsten were tested for comparison with the irradiated material. The tensile and yield strengths of tantalum were found to increase appreciably as a result of irradiation whereas the tensile properties of unirradiated Ta-W alloys prepared by arc melting showed that small additions of tungsten do not signicantly increase the strength of tantalum. These results indicate that the major part of the increase in strength resulting from irradiation of tantalum can be attributed to fast-neutron damage and that any contribution produced by the conversion of tantalum to tungsten is a minor one. (auth)
Date: November 18, 1960
Creator: Franklin, C. K.; Stahl, D.; Shober, F. R. & Dickerson, R. F.
Partner: UNT Libraries Government Documents Department

A Method of Correlating Irradiation Effects in Dispersion Fuels

Description: A method of correlating irradiation effects in dispersion fuels was proposed in which the effects of irradiation conditions are considered independently of material variables. Two simple failure models were devised (failure by creep and by short-term stress yield). Criteria which permit estimates of the relative severity of tests made under different test conditions but on identical specimens were developed. Numerical application of the procedures for 18-8 stainless steel with 25 and 30 wt.% UO/sub 2/ specimens was attempted. No positive verification of the short-term stress model was obtained, but the creep model yielded the approximate failure limits for both specimen compositions. (auth)
Date: January 20, 1960
Creator: Keller, D. L.; Hulbert, L. E. & Dunnington, B. W.
Partner: UNT Libraries Government Documents Department

Effects of Yttrium on the Fabrication and Tensile Properties of Two Modified Stainless Alloys

Description: Alloys containing 55 wt.% iron--22 wt.% nickel-- 17 wt.% chromium--2.5 wt.% molybdenum- 1.0 wt.% niobium-0.03 wt.% carbon-- 0.5 wt.% manganese-- 0.5 wt.% silicon with nominal additions of from 0 to 1.5 wt.% yttrium, and 36 wt.% iron-37 wt.% nickel--18 wt.% chromium--2.5 wt.% molvbdenum- 1.5 wt.% niobium-- 1.0 wt.% aluminum-0.05 wt.% carbon--0.5 wt.% manganese-- 0.5 wt.% silicon with nominal additions of from 0 to 2.0 wt.% yttrium, were prepared by vacuum- induction melting. Alloys containing 55 wt.% iron were successfully forged in air at 1900 deg F, rolled at 1850 deg F to 0.060-in. sheet and cold rolled to 0.015in. sheet. Fabrication of alloys containing 36 wt.% iron with more than 0.5 wt.% yttrium was unsuccessful. Addition of yttrium had relatively no effect on the yield and ultimate strength from room temperature to 1850 deg F. The ductility of fabricable alloys studied was increased at elevated temperatures by increasing yttrium contents. The greatest increase in ductility occurred at 1.5 wt.% yttrium. (auth)
Date: February 24, 1960
Creator: DeMastry, J. A.; Shober, F. R. & Dickerson, R. F.
Partner: UNT Libraries Government Documents Department

Solubility Limits of Yttrium and the Lanthanide Rare-Earth Elements in Chromium and Chromium-Iron Alloys

Description: The solubility limits of yttrium and the lanthanide rareearth elements in chromium and chromium-iron alloys were investigated at 2300 deg F and at room temperature. Bases of pure chromium and chromium alloyed with 10, 25, 50, and 75 wt.% iron were prepared with additions of cerium, dysprosium, erbium, gadolinium, holmium, lanthanum, lutetium, neodymium, praseodymium, samarium, terbium, thulium, ytterbium, and yttrium. Most of the elements investigated are estimated to be soluble to the extent of 0.1 wt.% or less in the base alloys. Dysprosium, erbium, and holmium exhibit solubility limits estimated to be between 0.1 and 0.3 wt.%. In general, the solubilities of these elements decreased slightly from 2300 deg F to room temperature. (auth)
Date: September 1, 1959
Creator: Epstein, S. G.; Bauer, A. A. & Dickerson, R. F.
Partner: UNT Libraries Government Documents Department

ALUMINA COATING OF UO$sup 2$ SHOT BY HYDROLYSIS OF ALUMINUM CHLORIDE VAPOR

Description: Uniform, dense coatings of alumina about 5 to 150 mu thick were applied to uranium dioxide particles 44 to 350 mu in diameter by hydrolysis of aluminum chloride vapor in a fluidized bed of the particles at 1830 deg F. The coated particles were resistant to nitric acid leaching, to oxidation in 1830 deg F air, and to thermal cycling from 6OO to 2500 deg F. After low neutron exposures, the coated particles showed excellent fission-gas retention at temperatures up to 2400 deg F in inert gas. Although not optimized in the study, the coating process appears to have commercial feasibility. (auth)
Date: October 25, 1960
Creator: Browning, M. F.; Veigel, N. D.; Cook, T. E.; Diethorn, W. S. & Blocher, J. M., Jr.
Partner: UNT Libraries Government Documents Department

Apparatus for the Study of Fission-Gas Release From Fuels During Postirradiation Heating at Temperatures Up to 1600 C

Description: An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)
Date: July 22, 1960
Creator: Barnes, R. H. & Sunderman, D. N.
Partner: UNT Libraries Government Documents Department

Canning Graphite for Gas-Cooled Reactors

Description: A preliminary investigation was made of techniques and materials for canning graphite to protect it for use at high temperatures in a nitrogen--oxygen atmosphere. Fabrication techniques for cladding bare and copper--plated graphite cores either in Type 316 stainless steel or Inconel X were developed. Specimens of the various combinations of core and cladding materials were subjected to simulatedservice conditions and evaluated. In all cases the Type 316 stainless steel-clad specimens failed by carburization and subsequent oxidation in relatively short periods of time. Although considerable trouble was experienced with rupture in the vicinity of the cladding welds during thermal cycling of the Inconel X-clad specimens, this material appeared to be satisfactory in other respects and is considered promising. A specimen of silicon-coated graphite eiad with Type 316 stainless steel was tested by heat treating for 624 hr at 1800 deg F. The silicon coating alloyed with the cladding material, formed a high-silicon diifusion zone, but prevented carburization of the stainless steel. (auth)
Date: January 1, 1959
Creator: Paprocki, S. J.; Carlson, R. J. & Bonnell, P. H.
Partner: UNT Libraries Government Documents Department

The Influence of Lubrication on the Compactability of Magnesium-Green Salt Blends for Bomb Reduction

Description: Lubrication of die surfaces with mineral oil or Dag 217 during final compacting of UF/sub 4/--Mg blends prevented seizing. Mineral oil application after every third press allowed 18 compacts before seizing became severe. Similar application of Dag 217 allowed 78 compacts. Mixing 0.33 wt.% Ceremul "C" with the powder allowed 40 compacts. Punch clearance had little effect on seizing. (T.R.H.)
Date: June 18, 1957
Creator: Paprocki, S. J.; Carlson, R. J. & Smith, E. G., Jr.
Partner: UNT Libraries Government Documents Department

Development of a Fuel-Element Leak-Detection System Based on the Principle of Isotopic Exchange

Description: The selective removal of halide fission products from an aqueous solution by exchange with the halide in a solid silver halide was studied as the basis for a fuel-element leak detector. The retention of fission-product halides on a silver halide column was investigated as a function of coolant flow rate, halide anion, and column size. Fission prcduct decontamination factors and predicted operating lifetimes were obtained for a number of reactor operating conditions. It is concluded that a sensitive, rapid leak detector for a water- cooled reactor could be constructed from a silver bromide or iodide column monitored by a neutron detector to detect delayed neutrons from the halide fission products. The feasibility of gross gamma monitoring was found to be dependent upon the intensity of the gamma background arising from absorbed fission products on the silver halide column. (auth)
Date: August 1, 1960
Creator: Howes, J. E., Jr.; Elleman, T. S. & Sunderman, D. N.
Partner: UNT Libraries Government Documents Department

Development of High-Strength Corrosion-Resistant Zirconium Alloys

Description: Approximately 100 ternary and quaternary spongezirconium alloys were screened for structural and cladding applications in a natural-uranium-fueled heavy-watermolerated power reactor. The alloy additions studied included2 to 4 wt.% Sn, 0.5 to 2 wt.% Mo, and 1 to 3 wt.% Nb. The effect of 0.1 wt.% Fe and 0.05 wt.% Ni additions to the experimental alloys was evaluated. All compositions were are melted, rolled at 850 ction prod- C from a helium- atmosphere furnace, vacuum annealed 4 hr at 700 ction prod- C, and furnace cooled. Room- and elevated-temperature hardness measurements were used to estimate the tensile strengths of the alloys, while corrosion resistance was evaluated by 1000-hr exposures to static 300 ction prod- C water. (auth)
Date: February 22, 1960
Creator: De Mastry, J. A.; Shober, F. R. & Dickerson, R. F.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY 1958

Description: Data are given on: creep propenties of 15% coldworked Zircaloy-2; the liquidus in Al-U; effect of chloride on corrosion of type 304 ELC stainless steel; corrosion testing of stainless steels, Ti 75A lead, and rigid PVC as container materials for HNO/sub 3/-recovery process; viscosities of MgF/sub 2/ and MgF/sub 2/-MgO slags; phase boundaries in the U-Zr-H system; oxidation of Nb, Nb-W, and Nb-V alloys in dry air at 1000 and 1200 deg C; nonoxidizing properties of UO/sub 2/- La/sub 2/O/sub 3/ and UO/sub 2/-Sc/sub 2/O/sub 3/; study of bounding fuedamentals; hydrogen adsorotion by Nb; diffusion of H in ZrH/sub 4/; reflector controlled heterogeneous boiling reactor; pressure bonding of Zircaloy- 2-clad compartmented UO/sub 2/ fuel plates for PWR; and evaluation of UC as reactor fuel. (For preceding period see BMI-1247.) (T.R.H.)
Date: February 1, 1958
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JANUARY, 1959

Description: Thermal-conductivity measurements are in progress on an unirradiated, unclad, natural U specimen. Data are presented on thermal conductivity measurements performed on UO/sub 2/. The creep properties of annealed and of 15% cold-worked Zircaloy-2 are being studied. A program was initiated to evaluate loss-of-coolant incidents in the PRTR by means of simulation on a digital computer. Research on the casting of hollow Al-35 wt. extrusion billets is reported. Further refinement of the method developed for the analysis of Mg in cement is in progress. The infrared and gaschromatography analysis of irradiated dodecane, decane, cetane. and octane, and their urea complexes, were continued. The manner in which U metal solidifies in cylindrical graphite molds is under study. Work has continued on development of a stabilized hightemperature nuclear fuel capable of operation in either oxidizing or reducing atmospheres. Progress in the stud of potential fueled moderators has continued with the determination of hydrogen-absorption isotherms for the Zr-25 wt. alloy. The effect of fast-neutron flux on the mechanical properties of AISI Tvpe 347 stainless steel are being determined and evaluated. The forging of Nb-U alloys is reported. Thorium-uranium alloys are being studied for the purpose of developing improved corrosion resistance and irradiation stability of the alloy by means of alloying and control of processing variables. The causes of fission-gas loss from refractory fuel materials is being investigated. Cermet fuel materials consisting of from 60 to 90 vol. % U0/sub 2/, UN, or UC dispersed in a stainless steel or Nb matrix are being investigated. The gas-pressure bonding technique is being investigated for cladding and bonding Nband Mo-base fuel elements and assemblies. Dispersion fuels consisting of UC and UN dispersed in stainless steel were irradiated in the WTR. Stress-cycling tests were continued on Inconel specimens at 1300 and 1500 F, cycled at 1 cps. The ...
Date: February 1, 1959
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JULY 1958

Description: Data are given on the thermal ccnductivity of Ti--6 wt. % Al--4 wt.% V. The creep strength of 15% cold-worked Zircaloy-2 is being determined for reactor components in the 290 to 400 deg C range. Nine serles of high-strength corrosion- resistant Zr alloys were prepared and corrosion tested. Results are included. The fabrication of Al-U alloys for uses in various reactors is presented. The development of a natural-U fuel alloy with improved corrosion resistance is presented. Results of corrosion tests on the natural--U alloy are presented. A study is presented to investigate the possibilities of applying a thin protectlve coating of Mo on the interior surfaces of the reactor system by electroless and vapor-plating techniques. An irradiation-damage program to detcrmine the extent of damage to type 347 stainless steel in iast-neutron fluxes in the core of the ETR is presented. The properties of the high-niobium portion of the Nb-U constitution diagram are presented. The development of Th-U alloys with increased irradiation stability and corrosion resistance is presented. Dispersion-type fuel specimens containing 24 wt.% fully enriched UC and UN dispersed in a l8 wt.% Cr14 wt.% Ni-2.5 wt.% Mo--balance Fe matrix and clad with type 318 stainless steel were fabricated for irradiation testing. Gas-pressure bonding is being investigated as a technique for the cladding and bonding of Nb and Mo fuel elements. Corrosion testing of Ti in Darex Process dissolver solution is presented. Studies were continued on ways of preventing stress- corrosion cracking of Carpenter 20 Cb in the Sulfex-Thorex system. A run is under way in a new molten-salt composition (62 M% NaF-38 M% ZrF/sub 4/). Specimens of INOR-1 and INOR-8, Hastelloy B, and type S-816 are being tested. A program is presented for the preparation of as-cast irradiation-test specimens of UC and the determination of some physical properties of ...
Date: August 1, 1958
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JUNE 1958

Description: Developments for Zirconium-Clad Fuel Elements. The thermal conductivity of Zircaloy-2 clad specimezs of U, U-l.5 wt. % Zr, and Zircaloy-2 was measured from 20 to 800 C prior to irradiation. Corrosion data for a 1000 hr test at 3O0 C in pure water of nine Zr alloys are given: Zr -Sn-Mo, Zr-Sn-Mo-Nb, Zr - Sn-Mo-Nb (trace Fe and Ni), Zr-Sn-Mo (trace Fe and Ni) Zr-Sn, and Zr-Sn (trace Fe and Ni). Developments for Aluminum-Clad Fanel Elements. Studies continued in casting Al and cladding with Al. Corrosion results are given for 15 min tests in 300 C water of U-Zr alloys with traces of V, Nb, Ni, Ti, Mo, Pt, Cr, Si, Al, and Ru in various combinations. Studies of Uranium and Uranium Alloy Fuels. Corrosion danta are given for a U-Mo-Nb alloy and some U-Mo-Nb-Ru alloys in 680 F water after 672 hours. Oxidation data are given for U-Nb alloys versus time at 350 and 400 C. Corrosion Problems Associated with Separation Processes. Studies of corrosion of alloys in Darex, Sulfex-Thorex, and fluoride volatility processes are reported. Data are given for INOR-1 and INOR-8 in NaF-ZrF/sub 4/ at 650 C for 500 and 1010 hours. Developments for SRE, OMRE, and OMR. Casting techniques for preparation of UC were studied, and post-irradiation evaluation of SRE Th-U specimens is reported. Studies of Na-Ta Compatibility at High Temperatures. These studies continued, data being given for creep of partially-annealed arc-cast Ta at 1200 F. Other work preparatory to developmental studies for PWR, MGCR, and NMSR is described. (For preceding period see BMI-1267.) (T.R.H.)
Date: July 1, 1958
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JUNE 1959

Description: 8 5 F 5 ; 9 6 9 7 4 / 1 < =15% cold-worked Zircaloy-2 at 290, 345, and 400 deg C is being continued. Research to identify factors affecting irradiation-induced volume changes in graphite by means of sink- float density measurements was oontinued. The program to simulate conditions after a postulated loss-of coolant incident within the PRTR was completed. Lapsed-time motion pictures are being made through a windowed autoclave of the corrosive action of high-temperature water on defected Zircaloy2 U specimens. Progress on the development of an isotopic-exchange leak-detection systems is summarized. A program to develop a thermal-neutron-flux munitoring system for the Hanford reactors is reported. A project is being conducted to determine the temperature and approximate composition of the ternary eutectic in the Al--U--Ni alloys. A feasibility study to determine if Ca coatings can be successfully put on Ni by arc-spraying methods is reported. Work was continued on the valence effects of oxide additions (CaO and La/sub 2/O/sub 3/) to UO/sub 2/. An investigation is being made of the effects of combined high pressure and temperature on UO/sub 2/. Postirradiation data are presented on fueled specimens of ZrH/sub 1.65/--2 wt.% U. An evaluation of the effect of irradiation on the mechanical properties of AISI Type 347 stainless steel is being conducted. Hothardness data are presented for high-purity Nb, Nb-- Zr, Nb-V, Nb-Ti, Nb-Fe, Nb-Cr, Nb-V-Ti, Nb-V-Mo, Nb-V-Fe, Nb-V-Cr, Nb-V-Al, Nb-Zr-V, Nb-ZrTi, Nb-Zr-Mo, Nb-Zr-Cr, and Nb-Zr-Fe. A summary is presented of corrosion results obtained on Nb, Nb-Zr, Nb-W, Nb-Mo, Nb-V, Nb-Fe, Nb-Ti, Nb-Ti-Cr, Nb-Ti-Mo, and Nb-Ti-V alloys exposed in high-temperature water and steam. Corrosion data are given for Nb-U alloys in air, CO/sub 2/, NaK, water, and steam. Techniques are being developed for the preparation and cladding of cermets containing 60 to 90 vol.% of fuel dispersed in ...
Date: July 1, 1959
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING JUNE 1962

Description: >Progress is reported on reactor materials and components, studies of fuels, general fuel-element development, preparation of UO/sub 2/ single crystals, radioisotope and radiation applications, materials development and evaluation, coated-particie fuel materials, corrosion studies of Fluoride Volatility Process, progress on USAEC/AECL cooperative program, fission-product deposition studies, radiationeffects study of candidate fuel materials for MGCR, gascooled reactor program, process development and evaluation of properties of F- 48 niobium alloy, and gas pressure bonding of beryllium-clad elements. (M.C.G.)
Date: July 1, 1962
Creator: Dayton, R.W. & Dickerson, R.F.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MARCH 1958

Description: The thermal conductivity and electrical resistivity of U encapsulated in Zircaloy-2 with NaK as a heat transfer material were determined prior to irradiation. The short-time rupture strength of 15% cold-worked Zircaloy-2 at 290, 345, and 4OO C is reported. Tests to develop the right-angle method of extrusion cladding of U with Al were continued. Pressure bonding of Al to Niplated U by temperature and gas-pressure tecchniques has produced sound metallurgical bonds. A series of 17 alloy compositions was prepared for corrosion testing in an effort to develop a natural-uranium fuel alloy with improved corrosion resistance. A stsdy of the oxidation of U0/sub 2/ was continued. The corrosion of stainless steel by chloride-contaminated nitric acid solutions is reported. Studies are reported on the solidification of unalloyed U ingots. A study of the reactions that may occur during the induction melting of U is reported. Specimens of enriched UN or UC dispersed in stainless steel and clad with stainless steel were investigated to determine tensile properties at elevated temperatures and resistance to irradiation damage. Electrical resistivity and tensile strength measurements were made on U0/sub 2/ stainless steel cermets at room temperatures. Work has continued on the investigation of hydrides of U-Zr alloys as fueled moderators. Investigations of the radiation stability of U-Zr alloys is presented. Corrosion data are reported on gamma- phase uranium alloys. The preparation of U-Nb alloys is described. Oxidation data on Nb and Nb alloys tested in dry air at 1000 and 12OO C are reported and discussed. Degassing experiments at 450 to 650 C were successfully performed on Nb-H/sub 2/, permitting calculation of diffusion coefficients. Diffusion coefficients for hydrogen in delta zirconium hydride are given. The use of Ti as a construction material continues to appear promising for both dissolver and the feed adjustment tank for the Darex ...
Date: April 1, 1958
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department