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The operation of the ORNL Volatility Pilot Plant for decontaminating and recovering uranium from molten-salt reactor fuels is discussed. A description of equipment, operating details, and performance of each system within the plant is contained. (C.J.G.)

The feasibility was studied and a cost estimate prepared of an experimental reactor to determine the operating characteristics of beryllia- moderated, gas-cooled systems wtthin a power limit of 10 Mw(t). The heat energy produced by the experimental reactor is to be dissipated in a heat dump. No machinery for production of power was to be provided. Other requirements were that the reactor should be capable of testing core types different from the current MGCR design, and the system should permit use of gases other than helium. It was further directed that the reactor should be designated BORE for Beryllium- Oxide Reactor Experiment. Reactor development work was mainly in connection with the BORE preliminary design. It was established that the most important information which could be provided by a 10 Mw(t) reactor experiment would be on performance of fuel elements and moderator bodies. This required that the experirment duplicate the power density in the fuel and moderator that would exist in the full size reactor and made it advisable to use full length fuel elements. This resulted in an unconventionally shaped core which is roughly cylindrical with the length more than twice its mean diameter. Studies continued on performance of fuel …

Progress is reported on fundamental metallurgy, longrange applied metaillurgy, and reactor metallurgy. Separate abstracts were prepared for each section. (W.L.H.) (W.L.H.)

Tables of temperature dependent coefficients are presented for the computation of thermodynamic properties of gaseous UF/sub 6/ using equations which express the properties as expansions in powers of pressure. The various coefficients listed were computed in one degree increments for a temperature range of 500 to 999 deg R. This set of tables was computed by use of a virial equation representation for the equation of state and the thermodynamic properties were derived from this representation. (auth)

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The concept of operating bearings and gears unlubricated in a helium environment was questioned as to fundamental soundness. Tests were made to determine if immediate gross failure occurs. A bearing of standard manufacture was operated in helium for 42 hr. A high degree of gradual wear was encountered. No gross or sudden failures were found. (W.L.H.)

An estimate is made of the illumination required for the 80-in. bubble chamber, based on the absolute sensitivity of the film. The required intensity is estimated to be 23.4 tube is calculated to be 145 cm/sup 2/. The results are compared with these obtained from 72- and 20-inch chambers. (D.L.C.)

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A manual is given of the equipment and procedures used in the Army Reactor (SM-1A) to control the water purity and makeup. In addition to a description of the primary purification control system, a discussion is presented of the water chemistry control procedures for the auxiliary systems (e.g., the spent-fuel pit, the shield tank, and the waste disposal system). (T.F.H.)

Ability of the originally designed SM-2 stainless steel fuel elements to withstand static transverse pressure gradients of normal reactor operetion was investigated. Fourteen simulated original SM-2 type fuel elements by Tungsten inert gas (TIG) and electric resistance welding processes were subjected to pressure gradients to 6 psi and measured for deflection during stress and distortion after stress. Tests provided a basis for selecting a reference SM-2 fuel-element design and established a method of predicting maximum plate deflections under anticipated SM-2 operating conditions. Results showed that TIG welded elements, slightly more prone to deflection, are preffered for SM-2 application because they are less susceptible to weld failure and that fuel plate thickness is the most significant parameter in determining plate strength. It was also found that equilizing the length of the outer and inner plates significantly reduces maximum plate deflection and that effects of weld spacing and side plate thickness were not significant within the pressure range of the tests. (auth)

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The possibility was reviewed that errors in calculations might have resulted in an erroneously high theoretical strontium activity which would explain the unexpected low activity found ion the strontium recovered from Purex wastes. The new, corrected calculations (49,213 curies/T) improves the accounting from 65 to 75% of theoretical. The discrepancy still should be investigated.

Variables affecting the texture and grain size of uranium during beta treatment are summarized. The study of the effect of time and temperature in the beta phase on the growth index (G3) and grain size of the final alpha product is tentatively believed to show that higher beta temperatures for short times (up to about seven minutes) tend to promote slightly more negative growth indices and that higher beta temperatures give rise to somewhat finer grain sizes. Results of studies of both Jominy end-quenched bars and several full-sized rods and tubes quenched by total immersion showed that large thermal gradients promoted negative growth indices and produced grains somewhat elongated in the direction of the thermal gradient. The effects of endcooling in full-sized pieces quenched by total immersion in cold water showed that the axial growth index is negative up to distances from the end of about half the wall thickness of tubes and about half the radial dimension of rods. The grain refinement penetrates to a lesser distance from the ends. In the radial direction the growth index for these same pieces is largely negative to a distance below the outer diameter of about midwall in two tubes studied. In …

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This report, for August 1960 from the Chemical Processing Department at HAPO, discusses the following: Production operation: Purex and Redox operation; finished products operation; maintenance; Financial operation; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

A mass flow measurement technique is described which has the capability of measuring homogeneous flow, slurries, highly corrosive fluids, and multiphase fluids. The device features ruggedness and reliability and has the ability to measure external to the flow. In the proposed system, the fluid is made to pass through a U-shaped tube wherein measurements of the augular momentum and density yield mass flow directly. As the fluid traverses the 180 deg bend, a radial force is generated, which can be measured with a force transducer. Density is determined by measurement of the radiation absorption in the fluid. (W.D.M.)

This report addresses the critical mass studies, part X.

With the exception of a brief period of slightly elevated chloride level in the secondary blowdown, water-chemistry conditions during the period were satisfactory. During the period, the reactor was shut down for end-of-core-life testing and rearrangement. A set of specifications covering all electronic and electromechanical mechanisms required to control the SM-1 reactor through the rod- drive motors and clutches was prepared and issued. Installation of instrumentation for plant response and system performance was virtually completed. Work on the interpretation of long-lived radiochemical data obtained at the SM-1 during core lifetime was continued. Analysis of all fissionproduct data collected during Core I life has started. Thirty-eight stationary and seven control subassemblies from SM-1 Core II were checked for alpha contamination by a gas-flow proportional-counting technique. The work on the final design of a waste-disposal system for SM-lA was stopped and an investigation of an interim system containing a bypass sampling system was undertaken. Work continued on tests 202, 203, and 204 in the activitybuildup phase of Test Series 200. Core- physics measurements were taken at end of Core I life to complete the series of measurements made throughout the lifetime of the core. (W.L.H.)

Progress on design studies under Task 3, the mechanical-design portion of the SM-2 core and vessel design, is summarized for the period Oct. 1955 to Mar. 1960. Task 3 covers the mechanical design of the reactor vessel, vessel closure, nozzle penetrations, steel reflector, core support structure, flow divider, control rods, absorbers, and fuel elements. Layouts showing the basic designs, major dimensions, and materials of construction are presented. Stresses for the reactor vessel selected were within ASME Code limits. The report does not contain final results of Task 3 work. (auth)

To determine the thermal stability of SM-2-welded plate type fuel elements, test specimens were subjected to temperature differences across plate width. Thermal deflections caused by the relatively cool side plates restraining the axial expansion of the fuel region were measured along the axial centerline of the test specimens. Region-averaged temperature differences varied from 0 to ll6/sup o/F, or about l35% of expected reactor operating differentials. Test specimens, machined from standard full-sized fuel elements, consisted of a single fuel plate and its proportionate share of element side plates, and displayed an l-shaped cross section. Thermal deflections of 0.005 in. maximum were measured at the expected reactor operating conditions of 87/sup o/F region- averaged temperature differences. With initial (cold) deflections assumed within the SM-2 tolerance of (?) 0.008 in., test results indicated that the total operating deflections will be (?) 0.013 in. maximum. (auth)

The Zirflex Process employs a boiling aqueous solution of ammonium fluoride and ammonium nitrate to dissolve zirconium or Zircaloy. Average unoxidized Zircaloy dissolution rates are from 10 to 15 mils/hr for the optimum charge solution of 5.5M NH/sub 4/F-0.5M NH/sub 4/NO/sub 3/ at a F/Zr mole ratio of 7. Zircaloy, which is oxidized by exposure to high-temperature air or water, dissolves at rates of threeto five-fold less. Cores of uranium, uranium- aluminum, and uranium dioxide are not severely attacked by the Zirflex decladding solutions. Only the soluble uranium enters the waste, with losses varying from 0.3 to 3.0 g/l. The Zirflex waste solution is neutralized to a pH of 10 before storage. This requires approximately 0.07 gallon of 50% caustic per gallon of decladding solution. The neutralized waste consists of nearly 20 vol.% of rapidly settling solids, which are easily slurried under turbulent flow conditions. These solids tend to settle out in streamline flow and therefore agitation is required during temporary storage. Conventional nitric acid core dissolution is generally applicable to Zircaloy-clad uranium and UO/sub 2/ elements since the core material is essentially free from zirconium. The addition of aluminum nitrate to the nitric acid dissolvent at an aluminum/ residual …

Preliminary design calculations were performed for a once-through type steam generator and reheater for advanced sodium power plants in the 300-Mwe range. Parameters and performance data are presented. (D.L.C.)

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