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Hydrostatic pressing was investigated as a method of fabricating long preforms made from uranium powder. A laboratory scale pressure vessel was constructed to evaluate the feasibility of this process. Various metal powders were loaded into flexible molds and subjected to pressure in an enclosed liquid. The high density of uranium caused the flexible molds to distort during filling and after pressing there was a tendency for the compact to adhere to the mold. Methods of minimizing these difficulties are suggested. Uranium powder, compacted in a plastic mold at 21 tsi had a density of 12.5 g/cc. During the course of the investigation, it was found that hydrostatic pressing of other metal powders presented advantages over conventional steel die methods, especially in the ability to press experimental shapes using economical equipment. Recommendations were made for further development work on both cold and hot hydrostatic pressing. (auth)

On July 27, 1954, 965 grams of plutonium alloy contained in three standard quart size ice cream cartons were being removed from the process line by two process operators using the plastic bag technique. Shortly after the plastic bag scaler had been energized a brown spot appeared on the plastic bag. The glowing turnings burned through the containers and plastic bag and continued to burn until completely oxidized. Personnel contamination was minor and readily removed. The contamination was confined to the room of the incident and was deposited in gross amounts in the area surrounding Task IV and V. This report includes a description of the incident and subsequent decontamination activities and a summary of plutonium recovery. Facts unknown about the incident and the reasons for these facts not being known are discussed. Plans for the future for reducing the possibility of recurrence of a similar incident are formulated.

The equations used in advanced reactor core calculations and a semigraphical method for their solution are presented. The following fuel elements are analyzed: continuous plates, parallel-flow rods, cross-flow rods, tubular flow channels in a solid matrix of fuel, and pellets. (M.H.R.)

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S>A convenient modified form of Maxwell distribution is chosen. The modified expressions for the energy transfer from the ions to the electrons in a plasma and the bremsstrahlung from the electrons are calculated. Using the expressions some possible steady-state conditions for the ion and electron gases are derived and compared with those for the usual Maxwell distributions. (anth)

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This document is a HAPO report dated September 20, 1957. At the time of this report, there had been many different methods suggested for improving the Purex organic recovery system. Some of these suggestions been tried in the plant and others had not. Information concerning this system and, particularly, methods for improving it were wide spread. At the time of this report was no immediate source of information available on the proposed and investigated methods for improving the Purex solvent recovery system.

This document details activities of the Irradiation Processing Department during the month of August, 1957. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and engineering; operations; production and reactor operations; facilities engineering operation; employee relations operation; and financial operation.

The distribution of H/sub 2/ among feed materials and products in the thermite reaction producing dingot U was studied. The H/sub 2/ content of Mg, molten U, and molten and solid MgF/sub 2/ density oxide to 98 to 99% for the high specific surface, ow bulk density oxides. The observed values for properties of the U/sub 3/O/sub 8/ samples correlated better with conversion of U/ sub 3/O/sub to UF/sub 4/ than did the observed values for properties of the corresponding UO/sub 2/ intermediates. (auth)

Tests of the HRT fuel and blanket recombiners at several gas concentrations and steam rates disclosed satisfactory efficiency at approximately design conditions. At lower gas concentrations and steam fiow rates a small recombiner downstream from the condenser is necessary to maintain 100% recombination. It is recommended that the platinized catalyst in the fuel and blanket HRT recombiners not be replaced, and that a small recombiner be installed in the 1-in. off-gas line from each recombiner condenser. (auth)

plants ln England and France. With the increasing de output of given designs and probably allow operation at higher polymer contents than orignally foreseen, thereby reducing the make-up requirements. The physical characteristics of the OMRE such as critical loading, temperature coefficient, and general stability appeared to be close to the predicted values. Radiation levels in the primary circuit area during full power operation appear to be so low that maintenance is possible during operation. The reactor has been run for a full month at 30% polymer concentration and is, at the time of this writing, brought to a still higher steady state percentage of breakdown products ln the coolant stream. No evidence whatsoever of fouling or precipitation has been observed. The reactor behaves in a routine manner in all respects and invites immediate application of the OMR principle to reactors for large scale ceniral stations. Final design on one 11.4 Mwe unit for the city of Piqua, Ohio, has now stanted. A short description is given of OMR power reactors. The use of magnetic jack mechanisms for control and safety rods provides a reactors top shield without penetrations, as well as an unpenetrated lower core vessel, still avoiding any interference …

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 …

Spatial patterns of radioactive contuamination on forest foliage were measured by gamma spectrometry and are discussed with respect to local vs. world- wide origin of the fallout and implications for ecology, health physics, and management of nuclear facilities. In September 1959, I/sup 131/ on dogwood leaf samples varied from over 500 mu mu c/g dry wt near Oak Ridge National Laboratory stacks to 1 to 7 mu mu c/g near the margins of the Oak Ridge Reservation. Stack fallout tended to occur closer to the source than was calculated from hourly wind data by an IBM 610 computer program based on Culkowski' s adaptation of the SuttonChamberlain theory of atmospheric diffusion and deposition. Over most of the Reservation levels of Ru/sup 106/ Cs/sup 137/ Zr/sup 9/ >s/sup 5/Nb/sup 95/ and Ce/sup 144/ were similar to levels found elsewhere (2 to 9, 1 to 3, 2 to 9, and 10 to 20 mu mu c/g respectively) and were presumably controlled by weapons fallout. Higher levels were found in small areas and indicate the need for attention to localized contamination, even though indirect estilevels considered hazardous from the standpoint of health physics. (auth)

The purpose of this report is to present some experimental data concerning boiling burnout and to discuss briefly the significance of these data in relation to the Hanford production reactors.

Activities are reported on development work on a mass flow system capable of measuring externally the properties of homogeneous flow, slurries, highly corrosive fluids, and multi-phase fiuids. In the proposed system, the fluid passes through an S-shaped tube wherein measurements of angular momentum and density yield mass flow directly. (B.O.G.)

An experimental evaluation of the shield design for the ML-1 mobile reactor was made at the Lid Tank Shielding Facility. Thermal-neutron fluxes, fast-neutron dose rates, and gamma-ray dose rates were measured behind slab mockups of the basic shield design and a number of possible variations. The designs embodied various combinations of lead, Hevimet, stainless steel, boral, water, and aqueous solutions of ammonium pentaborate at two concentrations. The after-shutdown decay characteristics of the basic design were determined, and data were obtained from a fairly accurate mockup of the stainless steel plenum and gas duct typical of the top and bottom regions of the ML-1 shield. Analysis of results and application to the final shield design are not reported. (auth)

In connection with an evaluation of the operability of a 36-inch diameter remote evaporator at the Idaho Chemical Processing Plant that was to be packed with a corrosionresistant neutron-poison packing for criticality control, an investigation in a 30-inch diameter vessel proved that air sparging effectively mixes solutions. The data showed that at similar spar;e rates the presence of the packing caused an increase in the time needed for complete mixing. The investigation showed that solutions are readily evaporated in spite of the presence of packing in the tank. (auth)