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

"Iodide-processed thorium foils were irradiated with 9-Mev protons at temperatures below -140 degrees C. the recover of electrical resistance upon annealing was studied in the range 0 degrees to 75 degrees where tempering curves showed rapid changes taking place. Determinations of the activation energy associated with this process were made and the mean value obtained was 1.22 ev. Correlations of this result have been made with those found previously for copper. From these comparisons, a tentative assignment of the motion of interstitial atoms in thorium has been made for this process. In addition, some evidence has been found which illustrates the corrosive action that water has on thorium at temperaturs as low as 0 degrees C."

"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.:

"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 …

"A method for solving multi-group reactor equations which arise in the diffusion approximation is outlined. Considerable work has been done on this problem at KAPL and ORNL. Their approach is to replace the differential equations by difference equations. Complications arise in this method where more than one slowing down medium is present since the fluxes are discontinuous at the interfaces. The primary purpose of this article is to develop an exact integral expression for the neutron flux which automatically satisfies the boundary conditions. An iterative method for obtaining the fluxes and critical neutron multiplication ratio based upon the above-mentioned integral expression is given. The only approximation used in obtaining the fluxes, in addition to the use of multi-group diffusion theory as the basic model, is the use of numerical integration to evaluate the analytic expression. The equations for a two region, two group spherical reactor are given in a form suitable for machine programing. The extension to more than two regions is also considered.

"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 …

"A study has been made of the distribution of tracer fission products and plutonium between small samples of molten uranium and solid uranium oxide, carbine, and nitride. The distribution showed the same behavior i general for all three materials: 1. The rare earth elements, Cs, Ba, and Sr were extracted primarily into the solid scrub phase. 2. Zirconium and Nb partially concentrated in the scrub phase. 3. Plutonium, Mo, and Ru tended to remain completely in the metal phase. The distribution of activities agreed with trends predicted from the thermodynamic data. Uranium oxide appeared to be the most desirable scrub material for removing large amounts of fission products from the uranium while leaving beind the Pu. In addition the uranium metal was not severley contaminated by dissolved oxide."

"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."

"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."

"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.

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"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."

"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."

"Heat production resulting from the absorption of gamma ray photons in thermal shields and the leakage of neutrons and photons from ferritic thermal shields are investigated. The gamma rays considered arise from three types of reactor radiation -- thermal neutrons, fast neutrons, and core and reflector gammas. The energy spectra of the fast neutron leakage and absorption have been investigated in some detail because of the significant contribution of fast neutrons to the heating of the concrete biological shield."

"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."

Introduction - The work of the Chemical Development Group has included studies on the thermal and radiation stability of organic materials suitable for reactor coolants, the thermal and radiation stability of zirconium hydride, reactor safety devices involving chemical systems, and general analytical development.

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"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.

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"Measurements of the intra-cell neutron flux distributing for a proposed Uranium Production Reactor have been made using a mock-up of a portion of the reactor core. Thermal neutron and thorium resonance neutron flux-distributions were investigated. As a result of the experimental measurements on the first mock-up, a decrease in thorium content appeared necessary in the reactor design studies. Experiments are now in progress on a second mock-up in which this change has been made."

"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.'