This document provides Part VII, Section I, Paragraphs 1 through 16 and Part VII, Section II of the Material and Equipment Section`s activities during the fabrication of reactor components and vessels at the New York Shipbuilding Corporation.

None

None

This document from Savannah River Plant reports progress in January 1953. Heavy water studies include corrosion inhibitors and sulfide deposition. Reactor materials studies include uranium fuel elements and aluminium sheaths and special materials to include thorium and bismuth. Reactor technology studies include control systems, tritium production, CMX, moderator purification, concrete shield heating. Separations technology studies include Purex process, 221 building equipment, and fuel element handling.

This report presents the details of the operations of the Savannah River Plant manufacturing division during the month of January 1955.

This document is the January 1956 Monthly Report of the Process Section/Manufacturing Division at the Savannah River Plants. It contains information on activities in each of the SRP areas (100-400 areas), as well as on utilities and general services.

None

It is desired at this time to present in summary form a factual history of the Building 234-5 operations, encompassing not only the immediate pre-start-up period from April to July, 1949, but including the major obstacles encountered in the succeeding interval to the end of 1949. This report is intended to serve not only as a review of 234-5 operations during this period but also as a documentary evidence of the difficulties experienced and the manner in which they impeded the start-up. This report presupposes a certain familiarity with the background of the 234-5 Building. To one unacquainted with the scope of this project, or for those desiring a background review, reference should be made to Document HM-253, ``234-5 Building Program Review``, by D.D. Streid, dated April 22, 1949. It must be realized that any attempt to discuss in detail all the circumstances and difficulties contributing to the start-up period would be disconcerting to the reader, and would very likely be repetitive. Therefore, although it is far from the writer`s intent that a cursory review will suffice, the presentation will be as nearly as practicable, chronological and comparative. For the sake of clarity it will be necessary to present several …

The General Engineering Laboratory has submitted a cost estimate and a proposal covering the design, fabrication, construction, and testing of a uranium slug canning machine. This machine will cover requirements for mechanization of only the operations occurring within the aluminum-silicon canning bath. These operations are: Can and cap preheating, Can and cap wetting, Canning assembly, and Quenching. Stated briefly, the machine is required to insert a prepared uranium slug into an aluminum can and close the can opening with an aluminum cap. All assembly operations will be carried out beneath a molten bath of standard Hanford aluminum-silicon bonding alloy. The uranium slug is preheated and prewetted before being manually transferred to the slug assembly and quench machine. The process performed by the machine will be integrated with the manual slug preheating and wetting equipment so that proper preheating and wetting is attained before the slug is inserted into the machine. After assembly the completed canned slug will be transferred to the water quenching station, where the components are held firmly together until the molten aluminum-silicon has frozen, forming a homogeneously bonded assembly. This report provides a detailed description of the machine.

This quarterly report details activities of the Metallurgy Research Sub-Section for the months of October, November, and December 1995.

The schedule for the installation of the PAR slurry loop experiment in the ·South Facility af the ORR has been reviewed and revised. Changed design philosophy to include maintainability of the loop auxiliaries required extension of the main construction period to the end of,June, with a three month period ending in September, for test and run in during which construction will be completed.

A review of the present design information and studies relating to nuclear power plants indicates that the weight of the power plant components, exclusive of crew shield and nacelles, may at a particular design point be represented by an equation of the form, W{subpp} = W{subo} + K{subt}T. By use of this, a power plant parameter psi is defined as the ratio of the part of the power plant plus nacelle weight that varies with thrust to the net thrust of the power plant less nacelle drag. The funamentall part that this parameter plays in determining the gross weight of a nuclaer powered aircraft is shown. This parameter depends on the design point Mach number and altitude, and the weight and drag increments attributable to the nacelle installation.

Minor changes in materials and methods resulted in increased yields in the production of fuel elements for the Bulk Shielding Facility. Attempts to bond uranium to a mercury-- silver alloy resulted in bonds of insufficient strength. Thorium alloys containing 2 to 4% Nb. Cr. Mn, Ti, Zr, Be. Al, and Si were prepared. The alloys containing Be, Al. or Si were hard, brittle, and not cold-workable. The alloys containing the other elements were coldworkable in the as-cast condition. Impact testing of pure thorium, using standard V-notch specimens, revealed a transition from brittle to tough behavior at 120 to 200 deg C. The determination of UAL/sub 4/: UAL/sub 3/ ratios, lattice parameters, and microstructures for the Al--Si--U system is reported. It was found that for a 20% uranium alloy, only 0.8% Si is necessary to completely suppress the formation of UAL/sub 4/. In general, ferritic iron--chromium alloys appeared more resistant to static corrosion by 1000 deg C lithium and lead than did austenitic alloys. Compatability tests with combinations of Mo, UO/sub 2/ BeO, Nb, and stainless steels 309 and 316 are discussed. True stress-natural strain-time diagrams for copper at room temperature are presented. (See also ORNL-2839.) (C.J.G.)

Heat transfer coefficients were determined for the eutectic mixture LiF- KF-NaF (Flinak) flowing in forced convection through circular tubes. Heat, electrically generated in the tube wall was transferred uniformly to the fluid during passage through small-diameter tubes of nickel, Inconel, and 316 stainless steel. The variables involved: Reynolds modulus (N/sub R//sub e/), 2300 to 9500; Prandtl modulus (N/sub P//sub r/, 1.6 to 4.0; average fluid temperatures, 980 to 1370 deg F; and heat flux, 9,000 to 192,000 Btu/hr-ft/sup 2/. Forced-convection heat transfer with Flinak can be represented by the general correlation for heat transfer with ordinary fluids (0.5 < N /sub P//sub r/< 100). The existence of an interfacial resistance in Flinak-Inconel systems was established and its composition determined. Preliminary measurements of thermal conductivity and thickness of film were made. The results verify the effect of the film on Flinak heat transfer in small-diameter Inconel tubes. Thermal entry lengths, determined from variations of local heat transfer coefficients in the entrance of the heated section, were correlated with the Peclet modulus. (auth)

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

Preliminary experiments indicate that there are no significant differences in the corrosion rates of zirconium or Zircaloy-2 after etching with the nitric--hydrofluoric solution or the sulfuric--nitrite solution, provided proper etching and washing techniques are followed. Incomplete removal of the residual etchant is deleterious to the corrosion resistance; however, this effect in the case of the sulfuric--nitrite solution is not as pronounced as in the case of the nitric--hydrofluoric acid solution. The anticipated advantages in the new etch were not completely realized. Additional development aimed at modifying the sulfuric--nitrite etch would have to be performed in order to overcome the disadvantages before recommendation for the adoption of the etch could be made. (auth)

Methods are described for metallographic preparation of binary U alloys having small percentages of Mo, Nb, or Si. Details are given of mounting, attack- polish technique, and electrolytic etching. Photornicrographs are shown of sintered specimens of U--Mo and U--Nb and of a cast specimen of U--Si. (auth)

Based on the results of detailed fabrication studies, an evaluation of the effects of varying the type and size of UO/sub 2/ particles, the type and size of stainless steel matrix powders, blending procedures, compacting pressures, sintering times, temperatures, and atmospheres, roll-clading temperatures and reduction rates, total cold reduction, and heat-treating times and temperatures was made for UO/sub 2/stainless steel dispersion fuel elements. Transverse tensile tests, creep-rupture tests, metallographic examination, radiography, density measurements, and x-raydiffraction studies were used to evaluate the structure and properties of the fuel elements. From these studies a reference fabrication procedure for GCRE fuel elements was established. The fuel element core contains minus 100 plus 200-mesh hydrothermal UO/sub 2/ dispersed in an 18-14-2.5 alloy matrix prepared from minus 325-mesh elemental iron, chromium, nickel, and molybdenum powders. Commercial Type 318 stainless steel is used for cladding. Core compacts are sintered in steps to 2300 deg F after cold compacting at 15 tsi. Evacuated picture-frame packs are hot rolled from a hydrogen muffle at 2200 deg F with a 40% reduction in thickness on the first pass and a 20% reduction in thickness on remaining passes. After annealing at 2300 deg F, the fuel elements are given a …

ABS>Published information is not available for accurate prediction of the electrical insulating characteristics of helium at high pressures and temperatures. In general the breakdown voltage increases as the gas pressure is increased and decreases as the gas temperature is increased. The relatively low breackdown voltage of helium accents the importance of additional investigation in this field. (auth)

Thorium nitrate, removed from irradiated Th metal processed in the Thorex pilot plant, was converted to the oxide and then to the fluoride in one pilot-plant-scale and two laboratory-scale runs. Activity distributions, decontamination factors, and safety of the process are treated. (D.L.C.)

To determine the safety of the array in the storage vault for the SM-2 experimental fuel plates, two criticality criteria were applied. A maximum of 18 fuel plates was stored in sthainless steel tubes and the tubes belted to a frame on the wall to prevent movement. No tube could go critical by itseIf. The vauit was then assumed completely flooded by water. In the first calculation, the fuel array was assumed to be distributed uniformly over the wall forming a large slab. This method indicated the array might be critical if the steel tube and cadmium lining were neglected. In the second method, a conservative calculation, wnich included the steel tube and cadmium lining was made. This method indicataed the array was subcritical. Calculations were then made of the criticalty of the SM-2 vault without the steel--cadmium tubes and wcoden blocks. The multiplication factor of the vault was also calculated. In order to determine the accuracy of these calculations, an ORNL critical experimental array was calculated applying the same analytical techniques. (M.C.G.)

An outline of the major problems which have appeared in reliability research at Sandia Corporation is presented and generalizations based on this experience are discussed. (J.R.D.)

Indenter gages, Wiancko gages, and interferometer gages were used to measure air overpressure vs time at essentially ground level stations for both the surface (S) and undprground (U) atomic explosions. For the S Burst several instruments were placed on a line extending from an overpressure region of 13 psi to a region of less than one psi. The air measurements for the U Burst ranged from 32 to 2 psi. (D. L.G.)

Experimental instrumentation was tested on Operation Upshot-Knothole in an endeavor to improve existing blast phenomena measuring equipment and techniques. Experimental designs tested include a field unit oscillatoramplifier using transistor circuit elements, a subminiature two-wire field unit, a commercially developed Vibrotron gage and amplifier unit, and a frequency deviation multiplier circuit for obtaining increased signal-tonoise ratios. The Wiancko pressure gage was also evaluated more fully. The results of the tests are described. Transistor circuitry was found to be little affected by atomic blast phenomena. (M.P.G.)