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The following report covers the study of a small composite stock at Oak Spring, Nevada, known as Climax granite to study the seismic effect of a deep nuclear explosion contained in a large volume of unfractured rock.

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To determine the effects of gamma radiation upon transducers, switches, and motors, two test assemblies were irradiated in the ANL Gamma Irradiation Facility at levels of 1.50 x 10/sup 5/to 2.10 x 10/sup 6/ rad/hr to a total dose of 1.0 x 10/sup 9/ rad. The components tested were a motor, a straingage pressure pickup, a quartz pressure pickup, and a high-temperature switch. The performances of the straingage pickup and snap switch were highly satisfactory over the test period. The quartz pickup, however, exhibited a sensitivity to radiation which rendered it ineffective for use in such an environment. Radiation damage to the motor was manifest as degradation of insulation, and evaporation and condensation of some of the products of degradation upon the walls of the test containment shell. As a result, the motor lost power, indicating a need for insulation more resistant to gamma radiation. Radistion damage to paints, varnish, grease, and conventional insu lations was noted, but the extent of this effect was not considered detrimental to the test. (auth)

The need for the first core for EBR-11 resulted in the design and construction of the Initial Fuel Loading Facility for this reactor. The plant was built to provide the required initial loading, to train personnel, and to test prototype equipment for the remote reprocessing of fuel materials in the EBR- II Fuel Cycle Facility. The facilities include: remotely manipulated melting, casting, and pin processing equipment, a degreaser, hoods and their atmospheric control system, a gas-purification system, fuelelement-assembly equipment, mold- preparation and balance room, bonding furnaces, a maintenance shop, and a change area. (auth)

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The development of swaging and vibratory compaction process for fabrication of clad UO2 fuel rods is described. The cost is less than 50% of that for an equivalent core fabricated from pellets. The irradiation testing of vibratory filled and swaged UO2 rods is reported for burn-ups from 2,000 to 12,000 Mwd/t UO2; the results indicate that the rods are capable of excellent performance power reactors.

"The quantum theory problem of constructing explicitly a local Lorentz invariant model field theory which has an asymptotic particle interpretation and gives rise to an S matrix different from unity is discussed. It is proved that if a local covariant field H(x) has a complete current, the S matrix associated with this field is identically equal to unity. Results show that in order to get an interaction, the current is not allowed to have a terminating expansion in terms of a free field. The absence of a finite connection between a free field and an interacting field tended to confirm the belief that in local quantum field theories with particle interpretation the requirement of the existence of scattering forces in the physical states are such that it is not pos sible to identify them with a linear space spanned by ""free'' particle states."

It is now generally known by those working on Tory II-A that the average maximum fuel element wall temperature obtained during the Intermediate Power Test was somewhat higher than the 2250 degrees F design value. The purpose of this report is to explain how this occurred.

This report talks about Analytical and preliminary design studies of nuclear rocket propulsion systems. Volume VI. Compilation of hydrogen properties,

A series of fires and explosions in U. S. Atomic Energy Commission facilities handling alpha -active materiais during the last five years resulted in reconsideration of safety problems associated with glove boxes and other equipment used to contain these materials. The literature on construction and operation of glove boxes for work with toxic inorganic materials not requiring biological shielding is reviewed as a contribution to this re-examination, with special emphasis on methods and equipment for working safely with plutonium and other alpha -active materials. An effort was made to point out the direction of current trends in this field. Detailed discussions of glove box designs and methods of experimentation in these enclosures are not included in this report but sufficient information is furnished for finding needed details in the referenced material. Methods for the detection and measurement of alpha -active materials and of impurities in controlled atmospheres are discussed. In addition, the literature on controlled atmosphere enclosures, glove boxes for non- toxic inorganic materials, and the technique of experimenting with such enclosures is reviewed. Some previously unpublished developments are reported. (auth)

A detailed description of the site, facilities, and characteristics of the UTR Test Reactor and operations to be performed are presented. Also included is an evaluation of safety considerations and an analysis of potential hazards which might result from installation or operation of this reactor. Based on satisfactory operation of similar prototype reactors in the past and on a detailed hazards analysis, it was concluded that the operation of the reactor at this site wiil not constitute a potential hazard to property or persons in the surrounding area. The application for construction permit and facility license is included. (M.C.G.)

This report addresses the configuration of Tatum Salt Dome, Lamar County, Mississippi.

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

Test work was completed on development of a stripping method for the amine extraction (Amex) process which produces a concentrated uranyl nitrate solution for shipment to the refinery. This procedure offers potential cost savings by simplifying the overall mill-refinery flowsheet. The process involves treatment of the amine extract with calcium nitrate solution to convert the uranium in the solvent to a nitrate complex, stripping the uranium with water or dilute nitric acid, and recovery of nitrate from the solvent for recycle by contact with a lime slurry.

One of the characteristics of the safety control system of the Hanford reactors which is of primary concern from the standpoint of reactor safety is the response time. The purpose of this study is to determine whether the safety control system is sufficiently fast-acting to prevent an intolerably large power excursion from developing. An intolerably large power excursion is one that would cause a melting of the uranium while the reactor power is above the previous operating level. The purpose of this report is to show the techniques used to analyze the speed of control of the Hanford reactors. This report will help to standardize the methods and reduce to a routine the solution of this type of problem. The analog circuits will be presented and explained. A report has been written previously on the speed of control. The changes in the problem have necessitated the redesigning of the analog circuit and also the issuing of this report. This report compiles and organizes the analog work done on the speed of control to date.

The objective of this supplement is to review the basis, justification and procedure so that this production test may be used permitting a determination of minimum downtime, when advisable for improving moderator coefficient data as function of exposure.

The resistsnce of the water --tributyl phosphate (TBP) interface to diffusion of uranyl nitrate was investigated by a photographic photometric technique. The resistance was below the limit of detectability in the experiments, even with a surface-active agent present in the system. The sensitivity of the experiments established an upper bound for the interfacial resistance to diffusion equivalent to that of a 25- mu film of TBP under steady- state diffusion. The results indicated that the rate-limiting step in solvent extraction of uranyl nitrate from water by TBP is transfer of the diffusing species between the interface and the bulk phase rather than across the interface. In the experiments a steady-state concentration profile was established across an interface by providing a source of uranyl nitrate on one side and a sink on the other in a special cell. Uranyl nitrate diffused along the concentration gradient and hence across the interface. A resistance to diffusion by the interface would be manifest as a departure from equilibrium of the uranyl concentrations immediately adjacent to and on either side of the interface. Accurate values of the interfacial concentrations at steady state were obtained by extrapolating the profile obtained from a few point measurements on …

Foreword: This report describes results of a series of evaluation and design studies conducted between July 1, 1960 and June 30, 1961, related to heavy water moderated power reactor plants.

A study was made of the neutron diffraction patterns obtained from Kr liquid under seventeen conditions of temperature and pressure at 117 to 210 deg K. The low temperatures were used because the diffraction patterns and the corresponding radial distribution functions are more sharply defined near the liqdid triple point. (J.R.D.)

This test authorizes installation of up to ten non-overbore size zirconium smooth-bore process tubes into KW Reactor and continued charging of these tubes with KVNS fuel elements until authorized by other means (Process Standards or PITA) or until this test is terminated. While the test authorizes up to 10 tubes, only two process tubes are immediately available and are all that will be planned for in the initial installation and charging. Goal exposure will be set at 800 MWD/T.

From introduction: This summary report of the geology of the Appalachian basin was prepared as part of the Radioactive Disposal Program being conducted by the USGS on behalf of the Division of Reactor Development of the AEC.

Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)

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