Introduction: This report describes areas of granitic rocks suitable for underground nuclear tests within Federally-controlled land in the continental limits of the United States. This information was requested of the U. S. Geological Survey by the Albuquerque Operations Office of the U. S. Atomic Energy Commission, and was compiled during March 1959 by D. C. Alvord, W. J. Carr, P. M. Hanshaw, S. P. Kanizay, C. S. Robinson, R. W. Schnabel, J. A. Sharps, and C. T. Wrucke.

Three edge loaded thoria-urania nuclear fuel samples were assembled into a capsule, irradiated and examined. This irradiation was the first of a series to develop thoria-urania fuel for high heat flux, specific power and burn-up operation in a sodium graphite reactor.

The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more …

This document presents preliminary chemical flowsheets for cladding and core dissolution of zircaloy-clad uranium dioxide and stainless steel-clad uranium-molybdenum non-production fuels. These preliminary flowsheets together with existing process flowsheets should be adequate for process development in the currently forecast non-production fuels dissolution facility.

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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)

The preliminary process design of a PRFR pilot leaching plant, which is proposed to be located in Cell B of Building ORNL, is considered. Chemical, physical, and nuclear parameters are investigated in order that the leaching operations may be carried out without any chemical or nuclear hazards.

A summary is presented of the basic technical iniormation that pertains to processes that are used at the Dana and Savannah River Plants for the production of heavy water. The manual is intended primarily for plant operating and technical personnel and was prepared to supplement and provide technical support for detailed operating procedures. Introductory sections contain some background information on the history, uses, available processes, and analytical procedures for heavy water. They also include a general comparison of the design and laserformance of the two plants and an analysis of their differences. The technology of the heavy water separation processes used, namely hydrogen sulfide exchange, distillation of water, and electrolysis is discussed in detail. The manufacture and storage of hydrogen sulfide gas and the process water treatment facilities are also discussed. (auth)

The objective of the test detailed in this report is to irradiate self-supported fuel elements under conditions of severity comparable in these expected for future loadings of this geometry, to attempt to determine the resistance to corrosion of cooled cladding, the effect of supports on cladding corrosion and coolant flow patterns, and the relative resistance to ``hot-spot`` type attack and rupture of ``projection`` fuel elements and rib supported elements. This test will authorize irradiation of four columns of self-supported and four columns of rib-supported I and E, 1.47% enriched fuel elements until two ruptures are sustained in each group on type demonstrates a significant factor of improvement in rupture resistance over the other.

8 5 F 5 ; 9 6 9 7 4 / 1 < =15% cold-worked Zircaloy-2 at 290, 345, and 400 deg C is being continued. Research to identify factors affecting irradiation-induced volume changes in graphite by means of sink- float density measurements was oontinued. The program to simulate conditions after a postulated loss-of coolant incident within the PRTR was completed. Lapsed-time motion pictures are being made through a windowed autoclave of the corrosive action of high-temperature water on defected Zircaloy2 U specimens. Progress on the development of an isotopic-exchange leak-detection systems is summarized. A program to develop a thermal-neutron-flux munitoring system for the Hanford reactors is reported. A project is being conducted to determine the temperature and approximate composition of the ternary eutectic in the Al--U--Ni alloys. A feasibility study to determine if Ca coatings can be successfully put on Ni by arc-spraying methods is reported. Work was continued on the valence effects of oxide additions (CaO and La/sub 2/O/sub 3/) to UO/sub 2/. An investigation is being made of the effects of combined high pressure and temperature on UO/sub 2/. Postirradiation data are presented on fueled specimens of ZrH/sub 1.65/--2 wt.% U. An evaluation of the effect of …

Sufficient changes in the original projection fuel testing schedule have occurred to make the original schedules confusing. It is the intent of this document to revise an up-date those schedules so as to be a more realistic guide for associated development programs.

At present, the Purex Plant employs a two-stage process for the recovery of nitric acid from high-level wastes. This report summarizes the results of a design study on the Phase I installation of a prototype demister (vapor filter) between the E-F6 Concentrator and the T-F5 Acid Absorber. Further, the results of an engineering study are presented on the feasibility of the subsequent Phase II bypassing of the E-F11 Concentrator and auxiliaries to permit single-stage acid recovery, once satisfactory performance of the Phase I prototype equipment has been demonstrated. 7 figs., 2 tabs.

The hazards associated with several possible mechanical failures were analyzed for the PRTR. The consequences of these failures were evaluated for inclusion in the Final Hazards Summary Report.

The Single Tube Prototype Mockup (STPM) was constructed to be used as a tool to evaluate the mechanical problems involved in operating and maintaining many components of the Plutonium Recycle Test Reactor (PRTR). This report has been written to acquaint interested HAPO components with the capability of the STPM and for use as an aid in scheduling tests and/or to properly evaluate testing results obtained from the mockup.

Since the calculation of the PRTR energy distribution was first carried out by J. R. Triplett, the design has become sufficiently fixed to allow a refinement of his values. The present analysis, also, includes a calculation of the fraction of energy which is released in the shroud and process tubers that flows to the primary coolant to the top and bottom shield coolant is taken into consideration. Nuclear data used in the original calculations still appears satisfactory and is, therefore, utilized in the present analysis.

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This report summarizes the research and development work carried out during March, April and May, 1959, for Budget Activity 2790 - Separations Development for Non-Production Reactors, The effort on Activity 2790 will enable Hanford to begin reprocessing in January, 1962, the fuel elements from power reactors which employ depleted or slightly enriched uranium fuels.

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The 234-5 Building ventilation exhaust is continuously sampled for the purpose of estimating the amount of radioactive (alpha emitting) material discharged to the atmosphere. Although a record is kept of the gross amount of radioactive material discharged, few data are available concerning the size and kind of active particles in the exhaust air. Knowledge of the particle size permits: (1) an estimate of the validity of samples drawn through the sampling system, (2) a better knowledge of what the active particle ground deposition pattern might be, and (3) may provide information relating to filter performance. The kind of radioactive material discharged is important in determining relative health hazards. The object of this work was to determine the size and kind of radioactive particles in the 234-5 Building ventilation exhaust. A secondary objective was to review present routine sampling of the stream with particular regard to the particulates to be sampled.

The rate of uranium sorption by a strong-base anion-exchange resin (Dovex 21K) from a uranyl sulfate solution (U 0.005 M, H2SO4 0.02M, SO4 0.2 M) was studied using a stirred vessel technique and measuring the U235 gamma radiation on each bead. Resin initially in the chloride form and the sulfate for was studied.

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Hydrogen-absorption isotherms were measured over the range 535 to 835 C for zirconium--1 wt.% and--25 wt.% uranium alloys. X-ray-diffraction studies were made over approximately the same temperature range for the zirconium--1, -- 25, and --50 wt.% uranium alloys. In general, the alloys resenable the zirconium- hydrogen system, modified by the presence of uranium. With 1 wt.% uranium, the phase boundaries of the zirconium--hydrogen system are shifted to slightly lower hy-drogen contents. With 25 wt.% uranium, the first - two-phase-' region shifts to a hydrogen content 20 wt.% greater than in the zirconium--hydrogen system, while the second cctwo-phase'' region is unchanged. The eutectoid temperature is increased from 547 to 601 C. Heats of solution of hydrogen in the alloys were found to range from --25.9 to --47.9 kcal per mole for the 1 wt.% alloy, and from --30.7 to --50.6 kcal per mole for the 25 in.% alloy-. The x-ray-diffraction data support the interpretation that, as hydrogen is absorbed, the alloys break down to form uranium and zirconium, and the latter absorbs the hydrogen. The entire ternary isotherms could not be deduced from the data. However, three aspects appear certains (1) the extent of the phase fields along the zirconium-- hydrogen …

Reaction rates of niobium with nitrogen were determined gravimetrically from 675. to 875 deg C with a recording microbalance and volumetrically from 1100 to 1600 deg C with a modified Sieverts apparatus. Diffusion coefficients and terminal solubilities were determined from 800 to 1600 deg C by the concentration- gradient technique. Tne reaction of nitrogen with niobium follows a parabolic rate law at 675 to 1600 deg C. The expression for the diffusion coefficient for nitrogen in niobium at 800 to 1600 deg C is given as well as the expression for the terminal solubility for nitrogen in niobium. (auth)

Developments in the production of various graphite products and data on the properties, outgassing, and radiation resistance of various graphites are reviewed. (C.J.G.)