"An experiment to determine the effect of reactor irradiation on the thermal conductivity of uranium-impregnated graphite at elevated temperatures as described. The results show a decrease in the thermal conductivity saturating at [approximately] 60 percent at a temperature of 700 degrees C; at [approximately] 50 percent at a temperature of 1000 degrees C; and at [approximately] 25 percent at a temperature of 1300 degrees C. It was found that after irradiation at a given temperature, exposure at a higher temperature resulted in an increase in the thermal conductivity. The converse was also observed. Within the precision of measurement there was no difference in effed between temperature changes produced by varying the fission rate in the samples and changes produced by varying the power in an external heater."

"Engineering was continued on the development of sodium cooled, graphite moderated type reactors. General studies were carried out as well as studies specifically devoted to the following: a. full scale poser-only plant, b. thirty-mega watt pilot plant, the SGR, c. sodium reactor experiment, the SRE. This work consisted of theoretical analysis of various aspects of nuclear performance; economic investigations of different fuel element, cooling system and plant arrangements; and experimental investigations related to the properties of certain materials and to the development of components. Preliminary consideration was given to alternative reactor arrangements employing liquid hydrocarbon moderators and high temperature coolants other than sodium. In addition to a summary of the general design features of the SRD, a program was prepared outlining the proposed use of this installation.

None

"A series of thermal neutron diffusion length measurements has been made on non-multiplying lattice of lead-cadmium alloy rods in D2O. One-inch diameter rods in square lattice spacing of 4, 9, 6, 9, and 12 inches were used. Excellent agreement was found between theoretical and experimental values of the diffusion length. The analysis o the diffusion length measurement required a correction for the epithermal neutrons entering the exponential tank. These epithermal neutrons provided a distributed source of thermal neutrons upon slowing down in the lattice."

"Work has continued on high temperature methods for processing irradiated uranium fuel. Additional results have been obtained with fused halide treatment, solid scavengers and direct Pu distillation. With fussed fluorides about 95 per cent of the Pu was removed from a uranium sample, while treatment of uranium with HC1 gas removed almost all the Pu and many fission products. treatment of molten uranium with uranium oxide removed a substantial fraction of the fission products without removing Pu. Uranium carbide treatment results were similar to the oxide but not as effective. A small scale distillation of Pu from uranium showed that Raoult's law is obeyed."

None

"The role of ionization in the phenomenon of 'radiation annealing' of graphite has been studied by using a 1-Mev electron beam. Changes in the c-axis of a sample with a Hanford irradiation of 460 mwd/ct were studied. Two thermal anneals of 4 hours each at 350 degrees C proved sufficient to complete the thermal annealing at this temperature. The samples were then irradiated for 7-1/2 hours at a temperature of 340 degrees C. The samples received an irradiation of 47 microampere-hours, equivalent to ionization to an exposure of 200 mwd/ct in a Hanford reactor. No changes were noted as a result of the electron bombardment. It is concluded that the ionization is ot of major importance in radiation annealing.

"A design study is presented for a sodium cooked, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam generator at 910 degrees F and is returned at 420 degrees F. Steam conditions at the turbine throttle are 600 psig and 825 degrees F. Cost of the complete reactor power plant, consisting of the three reactors, and on 150-megawatt turbogenerator, is estimated to be approximately $43,165,000."

"An investigation of the chemical effects of 1-Mev electrons on BrF3 at 25 degrees C has been carried out. Pressure measurements taken during the irradiation suggest the presence of Br2 and BrF5 as decomposition products and a fractional distillation of the irradiated liquid confirmed their presence. The extent of decomposition was determined both by fraction distillation and spectrophotometric methods. The radiation effect seemed to reach saturation when approximately 10 per cent of the BrF3 was destroyed. The exposure necessary for the decomposition products to reach a concentration of half the saturated value was calculated to be 2.7 microampere hours/cc BrF3 while the "G" value was found to be 1.5. A qualitative comparison of irradiation dosages from the Statiltron with that expected from spent fuels revealed that little decomposition of BrF3 reagent is to be expected from 1-say cooled Hanford fuel (in pile for 100 days) while in the case of 1-day cooled MTR type fuel (in pile for 12 days) a saturated effect might be realized in 1-3 hours. Since at most only 10 per cent of the BrF3 is destroyed it is concluded that BrF3, from a radiation resistance standpoint, is a suitable standpoint, is a suitable reagent for …

"The Atomic energy Commission has undertaken a development program to provide the technology needed for the evaluation and economic design of nuclear power plants. This program is to be carried out during the next five years at several national laboratories and industrial organizations. The Sodium Graphite Reactor (the SGR) is one of those to be investigated and experimentally tested as part of this 5-year effort. The program on the SGR is intended to expand our area of information covering sodium-graphite technology, experimentally demonstrate the feasibility of this reactor complex and extend its performance limits, and apply in information developed to designs suitable for the full-scale nuclear power plant. As a principal part of this program, a Sodium Reactor Experiment (the SRE) is to be constructed and operated; it will be the major experimental facility in which the performance of this reactor will be studied and new technological advances tested. This report continues an earlier series 2-7 in which previous work on the SGR and the SRE has been described. In this report, the progress on the program is described in two main sections. Section A is devoted to work relating to the general technology of Sodium Graphite Reactors, and to …

"Scale-up experiments on high temperature fuel recovery processes have included the dummy run phase on the handling of 1-kologram samples of molten, non-irradiated uranium in the hot cell. The next step involves the use of spent X-10 fuel slugs. Small scale experiments with X-10 uranium on the extaction of Pu with Mg show that as much as 80 per cent of the Pu can be removed in pone pass. Treatment of uranium with fused fluorides can remove at least 90 per cent of the Pu in one pass. Oxide scavenging with ZrO2 is very effective in removing rare earths.:

"General Chemistry investigations reported herein includes: (1) the Organic Coolant-Moderator Program, (2) investigations on zirconium hydride, and (3) analytical chemistry."

The following progress report describes the examination of elements removed in comparison to a previous study on the power facility reactor operations.

The following progress report describes the examination of elements removed in comparison to a previous study on the power facility reactor operations.

A report regarding the irradiation behavior of Unalloyed hypostoichiometric uranium carbide experiment AI 3-11 and review

Quarterly report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the first quarter of the 1966 fiscal year.

From abstract: Charts in this report show axial stress distribution over exterior bellows surfaces induced by bellows axial deflection, and by internal pressurization. The influence of root rings on stress distribution is presented graphically.

Introduction: This is the fourth report in a series of reports dealing in particular with a reactor safety device of a double diaphragm type.

Abstract: A device to detect the presence of hydrogen in sodium has been developed. Such a device, installed in a sodium heated steam generator, would signal the presence of water in the sodium resulting from a leak in the sodium-water barrier.

Abstract: This report, intended as a working document, contains analytic representations and analog models of the Hallam Nuclear Power Facility as used in studies of the Control and Protection Systems.

From abstract: This paper presents temperature-entropy and Mollier charts for sodium, and describes briefly the method used for their construction, based upon data from the literature.

From conclusions: There is a close enough correlation for most metals to give the chart some utility in estimating vapor pressure of metals.

From abstract: A conceptual design study has been performed for a sodium cooled, graphite moderated, thermal power-breeder reactor utilizing the Thorium-Uranium 233 breeding cycle. Several aspects of the design of the system are considered but no attempt has been made to supply all the details. It appears that the design presented is feasible and will allow the production of economic power as well as full utilization of thorium resources.

Abstract: The fast and thermal neutron leakage from the 30 megawatt SGR-P4 reactor has been studied by three independent methods.