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FIELD STUDY OF THE AgPO$sub 3$ GLASS PERSONNEL DOSIMETER (U.S. NAVY DT-60)

Description: >The U. S. Navy DT-60 dosimeter was evaluated under field type conditions as to its reproducibility and accuracy of gamma x-ray, thermal and fast neutron and mixed radiation responses. Gamma and x-ray responses from 25 to 600 rads at energies in oxcess of 200 kev were found to be accurate within 20% in 92% of 160 dosimeters examined. Because the DT-60 was found to have no detectable fast neutron response, effort was directed toward exaggerating the thermal neutron response to approximate the total neutron dose from a nuclear detonation and also toward eliminating any neutron response in a mixed neutron- gamma field. This was accomplished by various combinations of lithium, paraffin and cadmium shielding. Data indicate that the DT-60 dosimeter can be modified to approximate more closely doses from specific types of mixed radiations; however, this generally detracts from its value in approximating other types of mixed radiation doses. A paired system of dosimeters, in conjunction with one of the fast neutron dosimetry systems is proposed as being the most satisfactory ar rangement to approximate mixed radiation doses to personnel in the field. (auth)
Date: January 1, 1959
Creator: Ballinger, E.R. & Harris, P.S.
Partner: UNT Libraries Government Documents Department

IONIUM (THORIUM-230) FOR RADIOISOTOPE PREPARATION--STATUS REPORT

Description: The general prospects of several radioisotopes are reviewed; the special properties of U/sup 232/ and Th/sup 228/ are poi nted out; and ionium (Th/sup 230/ ) and protactinium target materials are discussed from the sthndpoint of availability and chemical separations processes required for the preparation of U/ sup 232/ and Th/sup 228/. Outlines are given for potential schem es for the separation of U/sup 232/ and Th/sup 228/ from uranium milling pr ocess waste streams and from the irradiation products of Th/sup 230/--Th/sup 232/ mixtures. The high heat generating rates of these potent alpha emitters make them especially suitable for primary consideration as heat sources for small thermoelectric generators. The exceptionally high alpha activity suggests their use in special neutron sources as Ra-Be sources, and they may have sufficiently high neutron generating rates to be in contention with some of the smaller research reactors and experimental neutron producers. (B.O.G.)
Date: December 15, 1959
Creator: Coppinger, E.A. & Rohrmann, C.A.
Partner: UNT Libraries Government Documents Department

THE PREPARATION OF URANIUM DIOXIDE FROM A MOLTEN SALT SOLUTION OF URANYL CHLORIDE

Description: Uranium oxides in a molten eutectic mixture of NaClKCl were chlorinated by bubbling chlorine gas through the mixture. The reaction product, uranyl chloride. was soluble in the molten salt. Although UO/sub 2/ was the most common oxide used, the reaction was similar in the other oxides. Phosgene and aluminum chloride were also used as chlorinating agents. A dense, crystalline precipitate of pure UO/sub 2/ was prepared by the reduction of the uranyl chloride contained in the molten salt solution. The reduction was accomplished by contacting the salt solution with any of several metals, by reaction with hydrogen or dry ammonia gas, or by electrolysis. Several kilograms of UO/sub 2/ were prepared by electrolysis using graphite electrodes. The physical properties of the material made it potentially useful as a ceramic fuel material. The initial high particle density of the "as-produced" UO/sub 2/ was considered of great potential advantage for adapting this process to the refabrication of irradiated UO/sub 2/ into recycle fuel elements. (M.C.G.)
Date: October 20, 1959
Creator: Lyon, W.L. & Voiland, E.E.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING MARCH 1959

Description: S>In the program on the use of radioisotopes for quslity control in the cement industry, experiments were performed on the determination of Ca by EDTA titration using a radiometric endpoint, on sulfur analysis by radiometric endpoint, and on sulfur analysis by rsdiometric titration with Ba/sup 140/ ae BaCl/sub 2/. Work on the survey and evaluation of various industrial processes which might utilize intrinsic radioisotopes in process control was completed. A study of the solidification of U cast in cylindrical graphite molds is being made. Studies aimed at reducing the amount of additive needed to stabilize UO/sub 2/ were continued. The irrsdiation history of a 2 wt.% U--Zr alloy hydride capsule is reported. Investigation of the changes in the mechanical properties of AISl Type 347 stainless steel as affected by exposure to fast neutron flux is being conducted. The evaluation of selected Nb-base alloys for possible service in pressurized-water reactors was continued. A program to investigate the friction and wear behavior of rubbing surfaces lubricated by high-temperature sedium was initiated. A study of the fabrication characteristics and the mechanical and physical properties of Nb-rich U alloys is being made. Thorium--uranium alloys are being investigated with the objective of improving corrosion resistance and irradiation stability by means of alloying and control of processing variables. A research program to develop a physical model and mathematical theory to describe the release of fission gases from sintered fuel materials is reported. Techniques for producing cements of 90% of theoretical density or better containing 60 to 90 vol.% fuel dispersed in a metal matrix are being investigated. A gaspressure-bending process is being investigated as a method of cladding of ceramic and cermet-type fuels with Mo and Nb. The effects of forming pressure and initial particle size on the perosity and grnin size of UC compacts sintered ...
Date: April 1, 1959
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING OCTOBER 1959

Description: The creep properties of 15% cold-worked Zircaloy-2, as evident from results being obtained, are superior to those of the annealed Zircaloy-2 at test temperatures of 290, 345, and 400 deg C. The development of a fuel-element leak detector was continued with studies to determine the effect of AgBr concentration and the flow rate of the I/sup 131/ solution -on the exchange between solid AgBr and I. Modification and improvement of a thermal-neutron-flux monitoring system for application to the Hanford reactors were continued. A program is reported for the development of corrosionresistant welding alloys for use with vacuum- melted lowcarbon Hastelloy F when used as a container material for spent fuel- element decladding solutions. Aluminumuranium alloys containing 0.5 to 3.0 wt.% Sn or Zr are being evaluated for possible use as reactor fuels. Additional activation analysis was performed on three samples of cement raw materials. Analyses for manganese and sodium were completed. Current work on the intrinsicradiotracer process-control application has dealt with the problem of Fe removal from refinery streams. An investigation is being conducted on the effects of oxide additions to UO/sub 2/, relative to slabillzation toward oxidation and volatilization. An irradiation-surveillance program concerned with the effects of irradiation on the mechanical properties of AISI Type 347 stainless steel is reported. Evaluation is being made of several Nb-base alloys which should possess better properties than a V--10 wt.% Ti--1 wt.% Nb alloy (an acceptable cladding material for the EBR). The evaluation of selected Nb-base alloys for sevice in pressurizedwater reactors was continued. An investigation of the creep properties of Zircaloy-2 during irradiation at elevated temperatures is being made. An investigation of the fabrication characteristics, mechanical and physical properties, and corrosion behavior in various media of Nb-- U alloys is being made to determine their applicability as reactor fuels. Thorium - ...
Date: November 1, 1959
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

PROGRESS RELATING TO CIVILIAN APPLICATIONS DURING SEPTEMBER 1959

Description: Tentative creep data are reported for annealed Zircaloy-2 sheet. A program directed toward the development of corrosion-resistant welding alloys for use with vacuum-melted low-carbon Hastelloy F to contain HAPO spent-fuel-element decladding solutions was initiatBe. Research to develop more satisfactory fuels from the Al--U system is reported. The development of a radiometric method for the determination of CaO in Port land cement was completed. In the study of radiation-induced nitration of hydrocarbons, a series of thermal and irradiation runs was completed in the liquid phase of the HNO/sub 3/-- cyclohexane determine the effects of ultra-high preasure and high temperature on uranium oxides and on the reactions of uranium oxides with mixed oxides. The irradiationBurveillance program was continued on type 347 stainless steel. Tensile data are reported for Nb-base alloys. A summary is reported of corrosion results obtained on Nb alloys exposed in high-temperature water and steam. The creep properties of Zircaloy-2 during irradiation at elevated temperatures are being investigated. Corrosion data are reported for Nb--U alloys exposed to high-temperature water and NaK. A program devoted to the determination of causes of fission-gas release in UO/sub 2/ is reported. Cermet and ceramic-type fuels are being clad with Mo and Nb by the gas-pressure-bonding technique. Data are reported on the densification of UO/sub 2/ by various pressure -bending conditions. Methods of producing dense UC pellets by powdermetallurgy methods are being investigated. Techniques for the production of high-quality cast shapes of UC are being developed. The rates of interdiffusion of U and C in the U-monocarbide-dicarbide system and the rates of selfdiffusion of U and C- in UC are being investigated. Hydrogen migration in Kr under the influence of a thermal gradient is being studied. Neutron-activation and in-pile experiments are being conducted to determine fission-gas retention and the effect on radiation on fueled-graphite spheres. ...
Date: October 1, 1959
Creator: Dayton, R.W. & Tipton, C.R. Jr.
Partner: UNT Libraries Government Documents Department

CHEMICAL TECHNOLOGY DIVISION, UNIT OPERATIONS SECTION MONTHLY PROGRESS REPORT FOR FEBRUARY 1959

Description: A gamma scintillation spectrometer was used to measure diffusivity of uranyl nitrate in water during preliminary capillary experiments. During Fluorox run FBR-22, 90.4% of the theoretical amount of UF/sub 6/ formed was collected in cold traps and chemical traps. Toroid tests of flame calcined mixed Th-U oxide showed low corrosion rates, small changes in particle size and a low solubilization of uranium, while denitration of uranyl nitrate in a fluidized bed resulted in particle growth with uniform layers of uranium oxide. A half-time of 30 min for uranium anion exchange was measured in differential bed studies of uranium sorption on Dowex 21K. The Darex Reference flowsheet operation resulted in chloride removal to less than 50 ppm in solvent extraction feed from APPR head- end treatment. Unirradiated prototype Consolidated Edison pins were dejacketed with 6 M H/sub 2/SO/sub 4/ with uranium losses to the dejacketing solution of approximately 0.2%. An optimum procedure was developed for clarifying large batches of solvent extraction feed by sand bed filtration. Sheared sections of stainless steel clad UO/sub 2/ were completely leached in onehalf the time required for equal lengths of stainless tubes containing uncrushed pellets. Abrasive disc wheel to metal removal ratios were measured at cutting rates from 10 to 60 in./min. Dissolution of Zircaloy-2 dummy fuel elements in an INOR-8 dissolver with the NaF-LiF salt system resulted in vessel wall corrosion rates of 1-2 mils/run (approximately 8 hr of HF exposure). At a heat generation rate of 60 Btu/hr/gal of solid wastes, the maximum temperature rise in a 0.75 ft radius infinite cylinder (k = 0.1 Btu/hr sq ft ction prod- F) was 1270 ction prod- F in soil, 1150 ction prod- F in rock, and 1020 ction prod- F in salt. (For preceding period see CF-59-1-74.) (auth)
Date: June 11, 1959
Creator: Bresee, J.C.; Haas, P.A.; Watson, C.D.; Whatley, M.E. & Horton, R.W.
Partner: UNT Libraries Government Documents Department

CHEMICAL TECHNOLOGY DIVISION, UNIT OPERATIONS SECTION MONTHLY PROGRESS REPORT FOR MAY 1959

Description: The measured diffusivity of uranyl nitrate in water at 25 ction prod- C was 0.7 x 10/sup -6/ cm/sup 2//sec with about 40% average deviation. A program was started to develop nonnuclear uses for depleted uranium. Two continuous DRUHM reaction runs were terminated due to erratic operation of the sodium metering system. In the second Fluorox run with crude UF/sub 4/ which lasted for 29 hr, a total material balance of 94.8% was obtained and 17.9% of the theoretical amount of UF/sub 6/ was collected in cold traps and chemical traps. Room temperature flow rate-pressure drop calibrations of a multiclone (thirteen 0.60-in. diam hydroclones in parallel) for installation with the HRT replacement circulating pump were completed. Mixed oxides of U : Th = 0.08 : 1 and all have low yield stresses of 0.02 to 0.05 lb/sq ft compared to 0.2 to 1.0 lb/sq ft for normal Th-U or Th oxides of 1.5 to 2.5 micron mean diameter. The rates of uranium anion exchange from solutions containing between 0.025 and 0.20 M sulfate were measured and apparent uranium diffusion coefficients between 1.2 x 10/sup -7/ cm/sup 2//sec and 1.6 x 10/sup -7/ cm/sup 2//sec were calculated. In bench scale studies, the Darex reference flowsheet was successfully applied to stainless steel-clad UO/sub 2/ fuels (Yankee Atomic) and to aluminumuranium foreign reactor fuels. The corrosion of titanium A-55 was measured in the vapor and liquid phases of a modified boiling Thorex dissolvent (13 M HNO/sub 3/, 0.04 M F/sup -/, 0.1 M H/sub 3/BO/sub 3/) containing 0.0, 0.5, and 1.0 M thorium from dissolved Consolidated Edison pellets and the maximum corrosion rate was 0.6 mils/ month. Siliceous filter cakes resulting from the filtration of Darex solvent extraction feed solutions through porous metal filter elements were easily washed to a uranium loss of ...
Date: August 25, 1959
Creator: Bresee, J C; Haas, P A; Horton, R W; Watson, C D & Whatley, M E
Partner: UNT Libraries Government Documents Department

THERMAL EXPANSION OF URANIUM DIOXIDE. Final Report

Description: The thermal expansions of commercial uranium dioxide specimens were measured up to the melting point. The linear expansion of dense, normal grain size UO/sub 2/ follows closely the equationi L = L/sub 0/(1 + 6.0 x 10/sup -6/t + 2.0 x 10/sup -9/t/sup 1.7 x 10/sup -12/t/sup 3/). An anomalous expansion was noted in the temperature range 1000 to 1500 deg C. Above 2500 deg C the rapid vaporization and crystal growth of UO/sub 2/ necessitate the application of heating techniques which provide rapid heating and quenching in order to obtain reliable data. The use of solar and arcmelting furnaces for this type of measurement is described. (auth)
Date: April 15, 1959
Creator: Halden, F.A.; Wohlers, H.C. & Reinhart, R.H.
Partner: UNT Libraries Government Documents Department

Description of Purex Plant Process

Description: A brief summary, with reference literature for details, of pertinent and important process flowsheet conditions which are in use in the Purex Process is presented. (auth)
Date: April 30, 1959
Creator: Irish, E. R.
Partner: UNT Libraries Government Documents Department

THE TRANSFORMATION BEHAVIOUR OF SOME DILUTE URANIUM ALLOYS

Description: Metallographic and dilatometric studies of the heat treatment behavior of technical uranium and certain alloys of interest for use in natural-uranium fuelled reactors are described. The micro-structure and secondary phases present in high purity and technical purity uranium are first discussed and the time- temperature-transformation data presented. Results on binary alloys comprising U/ Al, U/Cr, U/Fe, U/Mo, U/Nb, U/Ti, U/V, and U/Zr, containing up to about 2 at.% of the alloying element, are described. Cr, Fe, and Zr were found to be the most effective grain refining agents. Al and V were satisfactory under some conditions whereas Mo, Nb, and Ti were ineffective. The grain-size results together with the time-temperaturetransformation curves indicated that grain refinement is associated with depression of the temperature of transformation on heat treatment, the extent of this depression for a given cooling rate being dependent on the alloy content. In practice it was found that alloys such as U/ Zr and U/Fe responded to water quenching treatments, and optimum refinement of others (e.k., U/Cr) resulted from isothermal treatment at 500 to 600 deg C. (auth)
Date: October 31, 1959
Creator: Jepson, M.D.; Kehoe, R.D.; Nichols, R.W. & Slattery, G.F.
Partner: UNT Libraries Government Documents Department

URANIUM ALLOY POWDERS BY DIRECT REDUCTION OF OXIDES

Description: A process is outlined for the production of uranium alloy powders by co- reduction of mintures of uranium oxide and alloy element oxides. The reduction of mechanical mintures of the oxides of uranium and alloy element with calcium in a sealed reaction vessel is shown to produce powder wtth a variation in particle composition, although of consistert composition over various size fractions. The particular alloy systems which are considered are uranium--nickel, uranium-- chromium, uranium --molybdenum, and uranium--niobium. The uranium-molybdenum and uranium--niobium powders are single phase (metastable gamma), which is of consequence in the production of dimensionaHy stable nuclear fuels. Potential applications of some of these alloys are discussed. (auth)
Date: October 31, 1959
Creator: Myers, R.H. & Robins, R.G.
Partner: UNT Libraries Government Documents Department

MICROBAROGRAPH INSTRUMENTATION

Description: Microbarographs are used at the Nevada test site to measure acoustic energy at certain locations with respect to blasts. These data are used then to decide if weather conditions are favorable for a nuclear test. The microbarograph and its operation are described in detail. (T.R.H.)
Date: June 1, 1959
Creator: Hansen, H.E.
Partner: UNT Libraries Government Documents Department

PRFR PILOT LEACHING PLANT-PRELIMINARY PROCESS DESIGN

Description: The preliminary process design of a PRFR pilot leaching plant, the proposed location of which is in Cell B of Building 3026 at ORNL, is considered. Chemical, physical, and nuclear parameters are investigated to assure safe leaching operations. Nitric acid solvents are used for leaching the uranium and/ or thorium from the sheared spent fuel elements, and the dissolved fuel is sent through a shielded pipeline to the extraction plant for further processing. Recommended materials of construction are 304L stainless steel and 3O9SCb stainless steel, and maintenance is by direct procedures. (auth)
Date: July 23, 1959
Creator: McLain, H.A.
Partner: UNT Libraries Government Documents Department

RADIOACTIVITY IN SILT OF THE CLINCH AND TENNESSEE RIVERS

Description: Surveys of radioactivity in the Clinch and Tennessee rivers during 1954 through 1958 are summarized. It is concluded that no immediate hazard exists due to the reconcentration of radioactive materials in downstream bottom sediments, However, if the amount of radioactivity in the bottom sediment continues to increase for the next few years, it will be necessary to re-evaluate our present waste disposal policy in order to further restrict the release of ralioactive wastes to the Clinch River. The most probable effect of the radioactive sediment on industry would be an increased background counting rate if sand from the river bottom were used in making concrete for the construction of counting rooms of instrument laboratories. The problem ofthe radioactivity in solution in the river water would have to be considered before using the downstream water as process water in the manufacture of film emulsions or other photographic materials, (auth)
Date: December 1, 1959
Creator: Cottrell, W.D.
Partner: UNT Libraries Government Documents Department

BRIEF REVIEW OF HEAT TRANSFER PROBLEMS ENCOUNTERED IN THE PRODUCTION OF MAGNETIC FIELDS

Description: The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)
Date: May 25, 1959
Creator: Alexander, L G
Partner: UNT Libraries Government Documents Department

EXTRACTION OF URANIUM, MAGNESIUM, ZIRCONIUM, AND CERIUM FROM BISMUTH WITH A FUSED FLUORIDE SALT MIXTURE

Description: The extraction of uranium, magnesium, cerium, zirconium, and niobium from bismuth with a molten mixture of sodium fluoride and zirconium fluoride was demonstrated. Comparative rates of extraction were obtained. The effects of high concentrations of magnesium and of hydrogen fluoride sparging on the extraction process were investigated. Tracer studies demonstrated that exchange occurs between zirconium dissolved in the bismuth and zirconium in the fused salt. The applicability of the fused fluoride extraction step to the processing of the Liquid Metal Fuel Reactor'' solution fuel is discussed. (auth)
Date: June 1, 1959
Creator: Adams, M.D.; Fischer, J.; Meyer, R.J. & Phillips, N.D.
Partner: UNT Libraries Government Documents Department

CONTROL ASPECTS OF VERY HIGH FLUX RESEARCH REACTORS (thesis)

Description: Submitted to Rensselaer Polytechnic Inst. The computation of reactor characteristics over a fuel cycle in a way suitable to the investigation of the control aspects of the reactor problem is considered. The dynamic problem including as independent variables the neutron energy, the spaces and time was studied. An additional condttion of versatility in the method used to obtain the solutions is necessary to explore the control aspects of the problem. The normal possibilities of the analog computers were surveyed, and a method was found to solve the reactor dynamic problem. The classical approach of considering the neutron energy dependency in groups and the spatial dependency in reactor regions is used. Only cases of regular symmetry are considered, so that the reactor three dimensional configuration is reduced by analytical methods to the study involving one space coordinate. Time is considered as a continuous variable. The Mighty Mouse Deactor is simulated, and the analog results are compared against published data. With the spatial dependency represented by three core and three reflector regions, the fast and slow flux distributions are within 5% of the digital computer solution of the same problem. At the end of the fuel cycles the flux distribution is essentially that of the digital solution with the values lowered 5 to 10%. Compared with hand numerical computations made considering the same number of core regions, the analog results show agreement within 1%. Some fundamental aspects of the long term reactor dynamics are discussed, based on results from the simulator. Illustrative examples of power transients, shut-downs control by burnable poison and localized control rods are included and discussed. (auth)
Date: May 1, 1959
Creator: Vianna, A.C.D.B.
Partner: UNT Libraries Government Documents Department

THE USE OF INCONEL AS A HIGH TEMPERATURE, CORROSION RESISTANT, THERMAL NEUTRON FLUX MONITOR

Description: Inconel can be used for thermal neutron flux measurements by means of its cobalt impurity or its chromium constituent where conventional monitors are unsuitable. The use of cobalt should also be applicable to other nickel alloys. Discriminatory counting is required. (auth)
Date: March 16, 1959
Creator: Guss, D.E. & Leddicotte, G.W.
Partner: UNT Libraries Government Documents Department

TESTING OF DUMP VALVE TRIM IN THE VALVE TEST LOOP AND THE PERFORMANCE OF DUMP VALVES IN THE HRT

Description: The Homogeneous Reactor Test requires an emergency or dump valve for rapidly releasing the contents of the high-pressure systems to the low-pressure systems. In order to determine the most suitable trim material for this service, a loop was constructed and used to test a variety of materials under simulated reactor dump conditions. As a result of the tests both 347 and 17-4 PH stainless steels were recommended and used in the HRT dump valves. The HRT core dump valve has not been replaced since the beginning of critical operrtions. The blanket dump valve has been replaced twice since criticallty; once apparently as a result of mechanical binding and once following a dump. (auth)
Date: August 1, 1959
Creator: Hise, E.C.
Partner: UNT Libraries Government Documents Department

PROCESSING OF MOLTEN SALT POWER REACTOR FUEL

Description: ABS> Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF--BeF/sub 2/ salt as a solvent for UF/sub 4/ and ThF/sub 4/. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF- BeF/sub 2/ solvent salt which is highly enriched in Li/sup 7/. The recovered salt is decontaminated in the process from the major reactor poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation, partial freezing, and metal reduction, indicated that such methods may give some separation of the solvent salt from reactor poisons, but they do not appear to be sufficiently quantitative for a simple processing operation. Solubilities of LiF and BeF/sub 2/ in aqueous 70t0 100% HF are presented. The BeF/sub 2/ solubility is appreciably increased in the presence of water and large amounts of LiF. Salt solubilities of 150 g/liter are attainable. Tracer experiments indicate that rare earth solubilities, relative to LiF-- BeF/sub 2/ solvent salt solubility, increase from about 10/sup -4/ mole% in 98% HF to 0.003 mole% in 80% HF. Fluorination of uranium from LiF--BeF/sub 2/ salt was demonstrated. This appears feasible also for the recovery of the relatively small ccncentration of uranium produced in the LiF- BeF/sub 2/ThF/sub 4/ blanket. A proposed chemical flowsheet is presented on the basis of this exploratory work as applied to the semicontinuous processing of a 600 Mw power reactor. (auth)
Date: April 1, 1959
Creator: Campbell, D.O. & Cathers, G.I.
Partner: UNT Libraries Government Documents Department