"The composition of the solder for the solder plug has been set as the tin-silver eutectic. Final tests on this solder show that life expectancies much longer than 6 months are probable with the current design. The design of the heater tube to contain the solder plug has been settled. This consists of a copper tube impregnated with U235O2. Arrangements have been made to have test specimens fabricated by powder metallurgy techniques. The equipment for the MTR in-pile test of trigger element response times has been largely completed and tested. The design of the complete inner capsule for the BF3 safety element has been developed as well as the cladding technique. Mock-up elements have been tested in the Hanford test reactor to determine the control that may be obtained with elements of this type, although the analysis of the results has not been made. Prototype elements are also ready for testing in the test pile, except for loading with B10F3. Experiments have been designed and submitted for approval for production pile tests of prototype."

"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 …

"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 …

"Engineering pertinent to the development of the sodium-cooled, graphite-moderated type of reactor was continued. This included work on problems related to the zirconium canned moderator, low enrichment uranium fuel, sodium piping, secondary coolant system, shielding, and the control and safety elements. A large fraction of the work was devoted specifically to problems of the proposed Sodium Reactor Experiment (SRE) configuration. In this connection, an integrated effort was initiated to prepare a complete preliminary design of the SRE by an early date. In addition, two alternate sodium-graphite reactor configurations were studied. One was an intermediate size, 145 thermal megawatt, unit optimized for the production of low cost plutonium. The second was a low power 10 thermal megawatt intended for power production, but in which sodium circulation through the core was entirely dependent upon thermal convection."

"Scale-up work on high temperature fuel recovery processes has progressed to the point where the (high temperature) vacuum furnace for several operations to the hot cells has been completed and tested under operating conditions. Small scale experiments on high temperature methods for processing molten irradiated uranium fuel have been made with spent X-10 fuel slug pieces. The results of direct Pu evaporation, treatment with fused fluorides and oxide scavenging were every similar to those found with tracer experiments."

"For a central station reactor power plant of the sodium-graphite type, two designs have been investigated. The first operates as a converter using slightly enriched uranium fuel and produces 150 electrical megawatts. The second operates as a thermal breeder using a U233-Th alloy fuel and produces 300 electrical megawatts. Consideration has also been given to the problem associated with the design and operation of the Sodium Reactor Experiment. All work related to the plutonium plus power sodium-graphite pilot plant, which was undertaken at an earlier date, has been completed."

"Continued progress has been made with the high temperature decontamination processes for irradiated uranium fuel. The fused salt treatment of molten uranium has been extended to UCl3. Plutonium and rare earths were extracted into the UCl3 phase. Direct plutonium distillation from molten irradiated uranium has been scaled up to the hundred gram scale. Solid scavenging experiments using uranium oxide, uranium carbide, and uranium nitride in contact with molten uranium have indicated fission product removal. A scaled-up investigation of the separation and recover of uranium from an SIR type ceramic fuel using the volatile fluoride process has indicated the feasibility of this separation method. The effect of irradiation on the decomposition of BrF3 has been further studies in experiments using the NAA statitron.'

"A study has been made of methods to prepare a fluid containing 1 gram of thorium per milliliter. The methods considered were solutions of thorium salts, suspensions of dry solids in water, and collodial suspensions. Thorium oxide, oxalate, and fluoride were tried in conjunction with one or more surface actants, but it was not possible to attain the required thorium concentration. Thorium hydrosol, produced by peptization of thorium hydroxide and subsequent electrodialysis, gave the necessary concentration of 1 gram per milliliter. A solution of 0.5 gram per milliliter was found to be stable to electron irradiation and did not flocculate upon shaking or standing. Selected surface actans which might be used as protective colloids were found to be unstable to electron irradiation.

"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."

This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.

Purpose: Automatic recording equipment was designed for use with a high temperature Sykes experiment in which calorimetric measurements were to be made to temperatures approaching 2000* C. At such high temperatures, radiation becomes the dominant mechanism for heat transfer. The temperature differences which are used to determine the magnitude of this transfer no longer are directly proportional to it, but must be related by the Stefan-Boltzman law of radiation.

Abstract: Uranium monocarbide shows considerable potential for use as a fuel in high temperature, high power density, nuclear power reactors. As Atomics International is proposing its use in sodium graphite type reactors, it was necessary to develop a process for fabricaing sodium bonded uranium carbide fuel elements.

Introduction: This report has been prepared to document the extensive analytical work required to design the control system for the Sodium Components Test Installation (SCTI), constructed for the Atomic Energy Commission at the Atomics International field laboratory.

Abstract: A fuel development program has been conducted, to investigate the behavior of U-10 wt% Mo alloy at the fission rates and temperatures anticipated in Core I of the Hallam Nuclear Power Facility.

Abstract: One hundred gram quantities of uranium dioxide, irradiated to burnup as great as 21,000 Mwd/MTU and previously reprocessed by AIROX (Atomics International Reduction Oxidation), were refabricated into high density pellets and encapsulated for re-irradiation.

Abstract: Atomics International conducted a fuel irradiation program to investigate and define the behavior of U-10 Mo fuel at the temperatures and fission rates anticipated in Hallam Nuclear Power Facility (HNPF) Core I.

Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.

Introduction: The purpose of this experiment is to determine the loss of electrical power output due to operating many unequally-heated cesium diodes, connected in series and parallel circuits.

From abstract: The response of the reactor to forced oscillations has been investigated with the pile oscillator technique and the data are presently being analyzed.

Abstract: The following report covers the final design of the modular steam generators.

From summary: The full 140 element loading of the core was completed on October 10, 1962. At this point, critical operation was begun for operator training and post-critical testing purposes.

Introduction: A series of critical and subcritical experiments have been carried out as part of the Advanced Sodium Graphite Reactor (ASGR) Critical Experiments program to provide physics information which would be directly applicable to the AGSR concept.

Abstract: An analytical and experimental effort toward a better understanding of the mechanisms involved in nuclear power reactor loss-of-pressure accidents is presented.

Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.