"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."

None

A preliminary study has been made of a solid homogeneous reactor for stationary power plant application. The core consists of graphite spheres impregnated with uranium and thorium, and the coolant is bismuth. This concept possible offers advantages over other solid fuel reactor systems with respect to simplification of core structure, fuel fabrication and fuel handling, and reduction of fuel inventory external to the reactor. From the results of this preliminary study, it appears that the potential cost of electric power from this reactor is competitive with that from other reactor systems which have been proposed for the same application. The Po210 produced in the coolant presents a decontamination problem, but is also possibly a valuable by-producgt.

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."

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.

"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 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 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.

"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."

Abstract: The design and testing of experimental equipment for remotely joining stainless steel tubing by heliarc welding is described. This apparatus consists of a modified heliarc welding torch which is hydraulically controlled to maintain constant arc voltage. A suitable arc voltage sensing and control amplifier circuit was developed for this application.

"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."

Report describing technical progress on the design and construction of an Organic Moderated Reactor Experiment (OMRE), to be operated at the National Reactor Testing Station in Arco, Idaho. "This is the first report of the series and coves the period from the initiation of the project to July 31, 1956. Also included as an appendix to the report is a detailed description of the OMRE facility" (p. 3).

From abstract: Impact tests on the Mark III Hanford safety element indicate that the design is mechanically adequate and does not constitute a personnel hazard.

From abstract: The vapor pressure determination of solid ThF4 has been completed. Experiments to determine the heat of formation of ThF4 by studying the reaction of ThF4 with SiO2 are described.

Abstract: The data presented herein, in the form of graphs, can be used to obtain the value of this energy.

Report describing technical progress on the design and construction of an Organic Moderated Reactor Experiment (OMRE), to be operated at the National Reactor Testing Station in Arco, Idaho. "This is the second report of the series, and covers the period from August 1, 1956 through October 31, 1956" (p. 3).

The design of a waste-gas treatment system for organic moderated reactors requires a knowledge of reactor waste-gas composition, generation rate, and radioactivity. To obtain data on these variables a continuous stream analyzer was constructed to analyze the waste gas from the Organic Moderated Reactor Experiment (OMRE).

Abstract: A magnetic-jack control-safety rod is under development for the 45.5 thermal megawatt Organic Moderated Reactor. The rod is "unitized," i.e., the poison element, drive, position indicator, and shock absorber are contained in a compact assembly which is inserted in a regular fuel channel opening in the core. Tests to develop components capable of operating under these conditions are described and results are reported.

Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).

Introduction: A low-capacity, low-decontamination plant can be built, as part of a power reactor complex, to avoid long distance transfer of fuel to a high-capacity aqueous processing plant. Activation of such a complex, with the processing plant adjacent to the reactor it serves, could decrease the cost of the integrated fuel cycle. The study of this concept is a major objective of the Processing Refabrication Experiment (PRE).

Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.

Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.