"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 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: 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: This report is a description of a 200 channel digital recording system used to record high temperature strain gage outputs and associated temperatures.

Abstract: A conceptual reference design of a 318 Mw(e) Graphite Moderated Boiling and Superheating Reactor (GBSR) is described.

Abstract: The design and application of two computers for the HNPF protective system is discussed.

From introduction: This report provides industry, plant operators, and the scientific community with information covering the results of the performance analysis.

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.

Abstract: Cladding and fuel material processing prospects are reviewed, and fuel system possibilities for near-term (~1 mill/kwh) and long-range (<0.5 mil/kmh) fuel cycles are described.

"This report describes second-cycle postirradiation examination and AIROX reprocessing-refabricating of uranium dioxide irradiated to an additional 10,000 Mwd/MTU burnup."