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Description: The concentration gradients of uranyl ion in aqueous and organic solutions were analyzed by taking a macro photograph of the desired gradient by monochromatic (436 m mu ) light transmitted by the solution normal to the gradient in an appropriate diffusion cell. Two Druhm runs were terminated due to malfunction of the sodium metering system and the third run was terminated when the UF/sub 6/ nozzle ruptured. Calculations of particle temperature versus time relations for the flame denitration-calcination method of preparing metallic oxide from nitrate solutions indicate that the times required for heat transfer are controlled by the rate of radiant heat transfer to particle surfaces instead of by conductive heat transfer within the particles. A completed experimental study indicated that electrolysis in a cell with a mercury cathode and a platinum anode is a practical process for removing nickel from HRT fuel solution. The apparent diffusion coefficient of uranium loading on Dowex 21K was shown to be directly related to the resin size. An explosion of sufficient violence to blow apart the Pyrex pipe dissolver occurred during the fifth Darex dissolution of simulated SRE fuel probably from a rapid gas phase reaction between hydrogen and oxidizing gases such as NO/sub 2/. Materials handling flowsheets were completed for (A) decladding, washing, recanning and storing spent SRE uranium fuel slugs and (B) the shearing and leaching of stainless steel clad UO/sub 2/ and UO/sub 2/- ThO/sub 2/ fuels. A literature survey is being conducted dealing with reactor coolant and coolant loop contamination and decontamination. During run R-17 for calcination of evaporated Darex waste, the same as run R-16 which deformed the bottom calcination vessel except that one of the three added pressure probes was vibrated to keep it unplugged, the bottom of the calcination vessel did not deform, and there was ...
Date: December 31, 1959
Creator: Bresee, J.C.; Haas, P.A.; Horton, R.W.; Watson, C.D. & Whatley, M.E.

A Separated 1.17-Bev/c K- Meson Beam

Description: This report describes the design and testing of a 1.17-Bev/c separated K{sup -} beam designed in the fall of 1958 in connection with a 15-in. hydrogen bubble chamber experiment. At the target the K{sup -}/{pi}{sup -} ratio was 1/140. At the chamber, 4.0 K{sup -}-meson decay lengths from the target, and after two stages of electromagnetic separation, the K{sup -}/{pi}{sup -} ratio was 12.5, corresponding to a total pion suppression by a factor of about 10{sup 5}. The K{sup -} flux at the chamber was 0.87 K{sup -} mesons per 10{sup 10} protons impinging on the target.
Date: December 30, 1959
Creator: Eberhard, Phillippe; Good, Myron L. & Ticho, Harold K.


Description: An investigation was conducted into the possibility of alloy additions to Zircaloy-2 to diminish hydrogen absorption during aqueous corrosion. The nickel in Zircaloy-2 is believed to be the major constituent responsible for the relatively high hydrogen absorption. Additions of up to 0.5 wt.% antimony, arsenic, bismuth, or tellurium were selected on the basis of their known ability to poison the catalytic effects of nickel in hydrogenation reactions of other systems. Results of tests conducted for a total of 224 days in 600 and 680 deg F water and 750 deg F steam revealed no decrease in hydrogen absorption in modified Zircaloy-2 containing the aforementioned alloy additions. Hydrogen absorption increased when these alloying elements were present in the range of 0.1 to 0.2 wt.%. Corrosion resistance also decreased with alloy additions in these ranges. A 2-atm. partial pressure of hydrogen in the steam or above the water did not affect hydrogen absorption in the alloys appreciably. The hydrogen partial pressure did not affect time to transition in corrosion rates, but did appear to produce higher weight gains than degassed water. (auth)
Date: December 29, 1959
Creator: Berry, W.E.; White, E.L. & Fink, F.W.

Prototype Freeze Trap Test

Description: A performance evaluation was made of a prototype liquid cooled freeze trap with sodium at 350 and 1000 deg F. The sodium freeze-off function was adequate for all test conditions encountered. The freeze-off occurred satisfactorily with the larger clearance provided by a test modification to provide 0.030 eccentricity to the rotating shaft. Turning the freeze-trap handle was successful in opening the unit for gas venting when 350 deg F sodium was used. For a seal formed with 1000 deg F sodium, 16 turns of the trap handle gave no measurable gas venting at pressures up to 30 psi. Melting out the seal opened the vent satisfactorily. All the major problems encountered during the test were mechanical and associated with the rotating feature of the unit. (M.C.G.)
Date: December 29, 1959
Creator: Cygan, R.


Description: The most adverse power distribution was revised based on a comparison of PDQ calculations and measurements made during the SM-2 flexible experiments. A review of the basic nuclear data and calculational models employed in the SM-2 nuclear analysis was rnade. A comparison between initial reactivily, hot-to-cold reactivity change, and xenon reactivity with experiment was rnade. Based on a revised power distribution, the core flow requirement was reestimated to be 7800 gpm. Tentative designs of the core support and fuel element structure were prepared and evaluated for pressure drop and flow distribu-tion. The ETR and MTR irradiation programs are suramarized. The TIG process for welding elements is discussed. Specimens of Eu/sub 2/O/sub 3/ dispersions in stainless steel were autoclave tested. Static deflection messurements indicated that a fuel element with cold rolled plates will have a deflection aproximately 18% lower than annealed plates. measurement of plate collapse on two elements indicated possible collapse in the range 140 to 164% of rated flow. Flow distribution and pressure drop tests were made for several core support structure configurations. Mockup experiments on the SM-2 initial cold, clean and SM-2 mid-life cores were completed. Limited power distribution and flux distributions were performed in the clean mockup. The hot-to-cold reactivily change was measured by aluminum displacement as 90. The average B/sup 10/ and U/sup 235/ worth in the clean mockup was measured as 43 and 0.157 cents /g. The reactivity effect of replacing control rod fuel assemblies by stationary fuel elements was measured in the clean mochup. Stuck rod positions were measured in the mid-life mockup. (For preceding period see APAE-memo-223.) (W.D.W.)
Date: December 24, 1959


Description: Modified Zirflex process flowsheets were developed for recovering uranium from the newer power reactor fuel alloys after discharge from the reactor. The STR (1% U97% Zr-2% Sn) and EBWR Core-1 (93.5% U-5% Zr-1.5% Nb clad in Zircaloy-2) fuels are used as examples of low- and high-uranium fuels, respectively. A dissolvent of 6 M NH/sub 4/F yields a solution of zirconium and a precipitate of ammonium uranous fluoride. In one process, ammonium hydroxide is added to produce insoluble hydrous oxides of uranium, zirconium and niobium. The NH/sub 4/F-NH/sub 4/OH supernatant is removed by filtration, partially evaporated, and recycled as dissolvent. The uranium and zirconium oxides are dissolved in nitric acid to yield a solvent extraction feed solution of low fluoride content. In an alternative process nitric acid and aluminum nitrate are added to the ammonium fluoride fuel solution to oxidize U(IV) to soluble V(VI) and prepare a stable solution suitable for solvent extraction. Chromic acid is also added in the case of the STR fuel. In a variation of this flowsheet for the EBWR fuel, only- enough 6 M NH/sub 4/F is added to dissolve the cladding. Nitric acid and aluminum nitrite are then added io dissolve the core. Insoluble niobic oxide, which carries about 0.03% of the uranium from the EBWR fuel, is removed by filtration from the solvent extraction feed solutions in the EBWR flowsheets. (auth)
Date: December 22, 1959
Creator: Gens, T.A. & Baird, F.G.


Description: Thorium oxide is formed by the calcination of thorium oxalate precipitated under carefully controlled conditions. Material is produced with mean particle diameters of 1 to 5 mu . Some of the thorium oxide had uranium added to it by decomposing uranyl carbonate on the thorium oxide followed by calcination. Most of the oxides prepared were calcined to 1000 deg C or more and size classified to remove particles greater than 10 mu . The oxides were prepared in 150-lb batches, with a complete cycle requiring 24 hr. (auth)
Date: December 22, 1959
Creator: Johnsson, K.O. & Winget, R.H.


Description: The results are presented of a survey of calculated gamma-ray levels at many points on the surface of the operating floor of the containment building for the Enrico Fermi reactor. That portion of the floor surveyed lies directly above the equipment compartment. The calculations were made with the aid of an IBM-650 electronic computer. The main source of radioactivity which gives rise to gamma radiation above the floor is the radioactive sodium-24 in the primary coolant system. This system was considered to be completely filled with sodium, and activated to an equilibrium activity of 0.05 curies/cc, which corresponds to infinite reactor operation at 500 megawatts power. No fission product contamination was considered for these calculations. The operating floor is 5 feet thick and of concrete and steel. The results of the survey indicate that above the equipment compartment the surface dose on the operating floor will in no case exceed 0.9 mr/hr at the expected full operating power of 430 megawatts. Included as appendices are derivations and methods of corrections from one set of concrete and steel thicknesses to another. (auth)
Date: December 22, 1959
Creator: Chaltron, W.F. & Hungerford, H.E.

Preliminary Studies of Scavenging Systems Related to Radioactive Fallout. Letter Report No. 10 for October 1 to December 1, 1959

Description: Progress is reported in the development of scavenging systems for the collection of fall-out. Data are included from tests of two cyclone separators for the collection of air samples. Results are included from laboratory studies on the scavenging of aerosol particles by evaporating and condensing water droplets. (C.H.)
Date: December 18, 1959
Creator: Stockham, J. & Rosinski, J.


Description: 7 = 9 9 9 9 7 7 7 = 9 9 9 95 : > @ 9 ; 5 8 @ = K : . ighpurity Nb deformed by impact or slow compression at - 196 deg C. An apparent phase transformation was detected in high- purity Ga deformed at 4.2 deg K. The specific heat of the group IV-A metals and alloys of Zr-In and Zr-Sn were measured from 1.2 to 4.5 deg K. In the Zr-rich portion of the Zr-Ga phase diagram, the alpha / beta phase boundaries of Zr are depressed by additions of Ga and the beta phase decomposes by a eutectoid reaction. The Cd pressures of alpha - and beta - Zr alloys containing 1 to 11% Cd were measured between 1090 and 1325 deg K. Crystal structures of several unreported transition-metal fluorides, rare-earth hydrides and nitrides were determined. Progress in the study of phase transitions in beta -quenched Zr-Nb alloys aged below the eutectoid temperature is reported. A high-temperature investigation of the order-disorder phase transition of a Cu31 at.% Au alloy has revealed an intermediate periodic antiphase condition. A previously described x- raydiffraction technique for the measurement of the thickness and strain of thin oxide films was applied to a series of five Cu/sub 2/O films grown on Cu single crystals. A new x-raydiffraction method for measuring film thickness, based on the integrated intensities of the Bragg maxima, is shown to agree very well with the thickness as determined from the line-shape analysis. A determination was made of the influence which electrostatic interactions with neighboring ions have on the energy n yields pi transition in the nitrate ion. Some information on the behavior of solute species in dilute solutions of Bi in BiCl/sub 3/ was obtained from absorption spectra. Studies of the gaseous ...
Date: December 16, 1959

Analysis of Radiation From Hnpf Cold Traps and Primary Sodium Pumps During Removal and Shipping

Description: The expected maximum contamination of the HNPF cold traps and primary sodium pumps was determined along with the maximum dose rates from these components during removal and shipping. Suitable shielding for casks to be used in the removal operation and for shipping these components away from the reactor site is specified. Access to an unshielded cold trap is limited by high dose rates, i.e., 100 mr/hr at 120 ft, after 180 days decay time. A handling cask providing a radial shield of 3 in. of lead will provide adequate personnel protection for the removal operation, if 180 days decay time is allowed before the trap is removed. An additional 2.4 in. of lead is required for offsite shipment of the cask. This additional shielding can be added after the trap is removed from the reactor building. Dose rates from the cold trap after the shield plug is removed from the access hole are shown. If direct line-ofsight exposure is avoided, dose rates to personnel will be below 100 mr/hr at any position, and below 10 mr/hr at distances greater than 20 ft from the access hole. Dose rates from the cask during its travel away from the hole, will be below 100 mr/hr at distances from the cask greater than 10 ft and below 10 mr/hr at 35 ft, if the cask is raised no more than 3 in. from the floor during its travel. Remote, unshielded handling of a primary sodium pump is feasible, since dose rates would be 100 mr/hr at 28 ft and 10 mr/hr at 90 ft, after ten years of operation, and providing that 14 days decay time is allowed to eliminate activity from the Na/sup 24/ film clinging to the pump. Dose rates after only one year of operation would be lower by a ...
Date: December 15, 1959
Creator: Rhoades, W. A.

Hazards Analysis of the Organic Moderated Reactor Experiment

Description: Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.
Date: December 15, 1959
Creator: Williams, R. O. & Allen, W. O.


Description: Two designs of sodium cold traps for the HNPF have been subjected to full scale tests, Performance features that were investigated include oxide removal efficiency, oxide capacity, pressure drop characteristics, economizer effectiveness, and temperature profiles, Results indicate that both designs should perform satisfactorily in the Hallam plant, (auth)
Date: December 15, 1959
Creator: Cygan, R.

HRT Process Flowsheets--Revised Edition

Description: Revised HRT flowsheets are presented. These revisions cover such items as relocation of freezer units on the lines, corrections to the numbering of lines, valves or instruments, and the addition of a few lines in the service areas. The waste and vent system flowsheet was redrawn as two sheets. (C.J.G.)
Date: December 15, 1959
Creator: Robertson, R. C. & Jones, J. E.


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.

IRRADIATION EFFECTS ON THE SURFACE REACTIONS OF METALS. Summary Report for October 1, 1958 to November 1, 1959

Description: Weight increases during the oxidation of irradiated foils of pure copper were greater than for unirraaiated specimens. Enhanced reactivity appeared to be strongest in the thin-film region up to about 5 mu g/cm/sub 2/. Oxide film (Cu/ sub 2/O) thickness for both irradiated and unirradiated specimens was approximately 1200 A. Radiation did not affect the reduction of Cu/sub 2/O during the induction period (period in which the reduction proceeds very slowly or not at all). In later stages of the reduction process, a serious lack of reproducibility was observed. Radiation effects on films of Cu/sub 2/O formed by prior oxidation of the copper substrate decreased the kinetics of secondary oxidation. The secondary oxidation curve exhibited a large gap at the point of interrnption for irradiation. The development of an automatic recording microbalance of high sensitivity and a furnace for studies in reactor radiation fields is reported. Measurements were made of the electrode potentials of irradiated (5.5 x 10/sup 19/ neutrons cm/sup -2/) copper, aluminum, magnesium, and zirconium. Cell potentials were found to be dominated by the oxide films formed on the electrode surfaces. The results indicate that radiation does affect the local anode reaction potential. No significant difference between the rates of the dissolution reaction of cold-worked and annealed copper specimens in Fe(NO/sub 3/)/sub 3/ was observed. Results on irradiated specimens indicated that irradiation enhanced the reactivity of copper specimens on the order of 5%. Irradiated copper specimens developed etch pits at a slightly greater rate than unirradiated specimens in
Date: December 15, 1959
Creator: Carpenter, F.D. & White, J.L.