From Summary: "This report is not only the final report under contract #14-01-001-189 but also the first submitted for publication and general distribution on the subject of fresh water from saline water by the Struthers Wells - Scientific Design process."

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

A series of calculations are described which evaluated the heat transfer and fluid flow aspects of the Plutonium Recycle Test Reactor Pressurized Gas-Loop Facility.

Report describing a 300-Mw(e) Supercritical Pressure Power Reactor's facilities, physics, economics, and problems encountered during its development. Appendix begins on page 103.

This purpose of this report is to establish a means of determining the mean effective temperature at which any specimen of Al-Si had been annealed.

This report consists of the description, drawings, connections, and schematics of the various control elements that make up the control system of the Engineering Test Reactor Critical Facility (ETRC).

The Stationary Low Power Reactor No. 1 (SL-1) was a boiling water reactor demonstration plant for remote military bases. The plant had been operated by Combustion Engineering, Inc. from February, l959 until January 3, 1961 at which time a nuclear excursion rendered the plant inoperable, This report summarizes the of operations of maintenance, test, health and safety, and administrative activities for the year February, 1960 until the incident.

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 volume describes the final results from completed investigational programs intended to test the performance of various concrete materials exposed to weathering.

This report summarizes experimental work completed on Subcontract AT-147.

From foreword: This report contains the test results and analysis of test results of flow studies and reactor vessel internals. Tests were performed at, below, and above the PM-1 design flow rate to observe flow distribution and pressure drops within the reactor vessel, as well as the overall reactor vessel pressure drop.

From summary: The SNAP 7 program is being conducted for the purpose of developing four radioisotope-fueled thermo-electric power generation systems. An important phase of this program is the processing of Strontium-90 into heat sources for these systems. The current effort involves the design and engineering analysis of two of the four thermo-electric generators.

Introduction: The Laboratory follows the recommendations of the National Committee on Radiation Protection concerning permissible exposure limits for those working with radiation. A system of controls and regulations has been established which should ensure that no person at the Laboratory will receive more than the permissible exposure. This manual is intended to provide the minimum background necessary for intelligent cooperation with the regulations.

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.

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.

Fuel Fabrication and Evaluation. Nearly stoichiometric, low nitrogen- content UC was synthesized in the plutonium glove box atmosphere. PuC and PuC- UC preparation studies were initiated by the oxide-carbon reaction. UC with nickel additions lowered required sintering temperatures, improved density over control straight UC samples, but did not improve densities over those obtained previously with UC. Microprobe analysis of niobium and 21/4 Cr- 1 Mo compatibility samples tested 4,000 hr at 820 deg C, showed no uranium penetration. Irradiation Test. The W1-1 capsule containing two clad UC specimens continued operation to 11,000 MW-d/ton by the end of April. The W1-2 capsule was inserted in WTR, but had to be removed after a short time because of a radioactive gas leak. Plutonium Facilities. The facility at The Carborundum Company started operation with plutonium in March. Modifications and maintenance continue to be required and are being completed with minimum effect on experimental work schedules. The major modifications of the facility at NDA, the once-through nitrogen and helium systems, were completed and initial operation is satisfactory. A performance test was initiated prior to operations with plutonium. Design and construction of equipment for the furnace box are in progress. (auth)

For the purpose of evaluating the compatibility of graphite and INOR-8 in a dynamic fluoride fuel medium, INOR-8 Forced Convection Loop No. 9354-5 was operated 8850 hr. The loop operated at maximum temperature of 1300 deg F and circulated a fluoride fuel of the system LiF--BeF/sub 2/-- UF/sub 4/. Post-test examinations of the graphite and loop components revealed no apparent corrosion or carburization problems. (auth)

Activities in a program to obtain UC pellets of maximum density for irradiation testing are described. A study was made of the effects of the UC powder thermal history, sintering time, sintering temperature, and additives on the UC sinterability. Preliminary results indicate that the UC powder synthesized under the most severe conditions of temperature and time (1740 deg C, 107 min) was less sinterable than powders prepared at lower temperatures, or at the same temperature for shorter time. It was also concluded that although both temperature and time of sintering pellets affect density, temperature is more important. Additions of Fe (0.1 to 1.0%) improve sinterability and promote grain growth. The final density obtained on sintering in a vacuum was about the same as that resulting from sintering in He at atmospheric pressure. (J.R.D.)

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)

Activities are discussed for work done on the design, components development, and engineering analysis of the MSRE, and materials development studies including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing. (B.O.G.)

LAMPRE-L The reactor was started up and held st 160 deg C and at hot critical for various measurements. Some of the 160 deg C measurements included rod worth, temperature coefficient, and shim-down extrapolated mass. In the approach to hot critical, multiplication was measured during the temperature increase from 160 to 500 deg C and found to increase markedly when the core meltsd. Reactivity changes due to jarring of the fuel capsules to remove bubbles and voids in the melted fuel were observed. Zero measurements made at 48O deg C are reported. Power coefficient measurements at 0 to 46 kw and power demand experiments were made, and the 50-kw transfer function is plotted. Shielding inadequacies are discussed briefly. The results of a fuel pin scan are given. Observations on the performance of the sodium system are discuaaed. Type 17-4 PH stainless steel was found to undergo additional grain boundary precipitution during long-term exposure to temperatures below the hardening temperature, and measurements of its properties are reported. fuel and Container Development. Impurities in the LAaMPRE-1 fuel which has a corrosive effect on the fuel container are discussed. Observations in the Pu -Ce-Co phase diagram are reported. Different types of LAMPRE-1 capsule …

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.)

A preliminary chemical flowsheet was developed on a laboratory scale for the preparation of feed and two solvent extraction cycles for the processing of Si-containing Al--U alloy, 20% enriched, fuel elements. Major process steps include dissolution of the fuel assembly in Hg-catalyzed ritric acid, removal of silica by coagulation during feed clarification, and recovery of U and pu by extraction with a tributyl phosphate solvent and selective stripping. Criticality control in existing process equipment is maintained by the use of internal neutron poisons and concentration control throughout the process. (auth)

Methods and techniques for measuring the conductances of reactor solutions at elevated temperatures and to interpret the results with electrolytic solution theories are discussed. Criteria applied to room-temperature conductances for determining complete ionization of 1-1 electrolytes were applied to the Noyes high-temperature conductance data obtdined on NaCl, HCl, KCl, NaOH, AgNO3, and H/sub 3/PO/sub 4/ solutions. lt was found that the first four electrolytes behave as strong electrolytes up through temperatures of 218, 218, 281, and 156 deg C, respectively. Application of the Bjerrum theory showed that, somewhat above each of these temperatures, these electrolytes begin to associate by ion-pair formation. H/sub 3/PO/sub 4/ solutions associated at room temperature and above. AgNO/sub 3/ solutions appeared to asscciate at all temperatures. The equivalent conductances and infinite dilution and the dissociation constants for each of the electrolytes were calculatsd. An apparatus was constructed and measurements were made on the conductances of LiOH and NH/sub 4/OH solutions over the temperature range of 100 to 520 deg F. Equivalent conductances at infinite dilution were calculated. The pH of reactor coolants using LiOH or NH/sub 4/OH was calculated from the disscciation constants for water, LiOH, and NH/sub 4/OH at elevated temperatures. (auth)