The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GCBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas cooled graphite-moderated (GGBR), and deuterium-moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site.

This report describes the preprocessing stage to be added to the present ORACLE-Algol translator which will enable it to translate programs containing procedures. The Algol 60 procedure is an extremely flexible tool, and its full implementation presents a number of difficulties. Since it is undesirable to undertake major revisions of the existing parts of the translator and impossible to solve all of these difficulties in a processing stage, it was necessary to impose certain restrictions on the use of procedures. The result, however, retains the more basic features of the Algol 60 procedure concept. Most of the algorithms published in the Communications of the Association for Computing Machinery can be handled by the preprocessor with few minor or no changes.

Sanderson & Porter have carried out a series of studies over the last four years which indicate that the pebble-bed reactor way be an attractive way to obtain low-cost power. At the request of the Atomic Energy Commission, two design studies have been carried out on this concept at the Oak Ridge National Laboratory. The first of these a preliminary design of a 10-Mw(t) reactor experiment, the PRRE, was initiated September 10, and a report on the study was issued November 1960. The second phase of the work, a conceptual design study of a 330-Mw (e) central station, was initiated November 1, and is the subject of this report.

The objectives of this study have been to examine the problems of the pebble-bed reactor concept and to conceive a design of a facility for investigating the feasibility of this type of reactor. The design must provide for adequate leaktightness of the contaminated-gas system and adequate maintenance of contaminated components, the most vital feasibility questions of the concept.

Control rod materials, designs, and fabrication processes are compared for their relative economies. Control rod lifetime data are calculated with a simple approximation of nuclear worth depreciation. These data are used in conjunction with the estimated fabrication costs to determine the cost of using several absorber materials in a typical power reactor on a cost-per-year basis. The effect the control system has on core power density and fuel lifetime is included.

Previously reported results on the testing of small samples of boron carbide dispersed in nickel by electrolytic codeposition were adequately encouraging to lead to the development of a prototype control rod for operation in the Vallecitos Boiling Water Reactor. The operation of the control rod has been entirely satisfactory.

This report contains a listing of the bound-proton kernels for neutron scattering which were calculated in conjunction with the USAEC Control Rod Materials Program.

The physics calculations to be performed on the Consumers On-Line Process Computer are reported. As a result of these calculations, a three-dimensional power shape is simulated, a reactivity balance is approximated, and the neutron exposure of each third of fuel bundle and control rod are accumulated. The work was reported as a part of the Consumers Research and Development Program.

The objective is to establish, from test results, the relationship between pressure loss and orifice configuration, mass rate, and steam quality, for a Fuel Cycle fuel assembly.

Use of the Smith chart in studying stub tuners is demonstrated. Several properties of the tuners are shown, and charts for various wave length spacings, tuning trajectories, and admittances are included.

Technique used in deriving values of the linac ball tuner settings to produce various fields is discussed briefly, and tables showing ball tuner change numbers and values used in making up the ball tuner change numbers are given.

This report describes the design procedure for a reflectometer or directional coupler to replace the probe part of the D07-1020 assembly used at the Brookhaven Alternating Gradient Synchrotron linac. Mathematical treatment and diagrams of various reflectometers are included.

Some numerical calculations have been done which permit the deduction of many properties of the synchrotron oscillation problem. For the presentation of the numerical results, it is useful to review the formulation framework to which the results are to be fitted. These are worked out for an arbitrary circular accelerator. The algebra woud have to be modified slightly to include straight sections but the numerical results are still useful.

Graphs have been made which show the nuclearly safe parameters for uranium rods in light water with uranium enrichments up to five weight percent U-235. These data were to serve as a guide to those persons who may be involved with the maintenance of nuclear safety in handling and processing operations with slightly enriched uranium fuel elements. The data are applicable to fuel element fabrication and processing operations, and in general to those operations involving the handling and storage of fuel elements apart from reactors.

This report summarizes the research and development work carried out during December, 1959, and January and February, 1960, for Budget Activity 2790 - Separations Development for Non-Production Reactors. The major effort on Activity 2790 has been completed. Current efforts on the remaining problem areas will enable Hanford to begin reprocessing fuel elements from power reactors which employ depleted or slightly enriched uranium fuels in July, 1962.

Critical mass experiments to be conducted at HAPO will require dilute and concentrated plutonium solutions. Since very high plutonium concentrations (~2,000 g/l) cannot be attained in pumpable solutions, solid plutonium compounds will be utilized for this phase of the work. Mixtures of plutonium oxide and polyethylene, methyl methacrylate or paraffin compacted in containers have been considers for use in highly-concentrated plutonium systems of known hydrogen to plutonium atomic ratio (H/Pu).

A 19-rod Zircaloy-clad half-length PRTR spike element successfully irradiated to high plutonium burnout at full power in the ETR shoved partial bonding of the core and cladding. Unsatisfactory autoclave films on the Zircaloy cladding of the PRTR Al-Pu fuel elements are delaying final assembly of the first 30 clusters. Further conditioning of the autoclaves and the availability of permanent etching facilities are expected to correct the present difficulties.

The evaluation of the inductance of an electrical component is often important in understanding and operating an electrical circuit. Inductance is defined by several equations, the most common being e= -L d1/dt . This equation defines inductance as the constant which relates the back emf produced in an element to the time-rate-of-change of current.

Investigation of aqueous and steam corrosion at elevated temperature has been carried on at Hanford for several tears. A number of reports have been published on various phases of the corrosion program but, except for monthly reports, little has been written on the progress of the Corrosion and Coatings Operation alloy development program. This interim report will outline the more important current phases of the aluminum alloy program.

In the continuing search for better methods of effecting the secondary fusion welded closure in Alsi bonded aluminum clad uranium metal fuel element, an article on "The Twin-Argon Welding Process" by J. A. Donellon was recently found in the September 1954 issue if the British Welding Journal. From a review of the article, the process appeared to have enough possibilities to warrant exploring it. Communication were established with the General Electric Company, Ltd..

The measurements reported here are companion measurements to those reported earlier in HW-63585. The only significant difference between the measurements is that 1.8 w/o enrichment UO2 fuel was used for the first set, and 2.6 w/o enrichment UO2 fuel was used for the measurements described in this report. The new results will be presented graphically, and for completeness, the details of the measurement will be included here as well as in HW-63585.

The development of a miniature probe was desired for measuring approximate single nuclide beta dose rate in solution and in various animal organs. This probe designed for biological experiments, was to have maximum possible sensitivity to detect low levels of nuclide concentrations. The desired dimensions of the light pipe were to be approximately one-fourth-inch diameter with lengths of three to twelve inches.

A brief series of experiments was performed to study the high temperature reaction of the Zircaloy-uranium-aluminum system at the request of Reactor Engineering Operation, Irradiation Processing Department. Knowledge of the high temperature behavior of this combination of materials is important to the evaluation of reactor behavior during possible accidental over-heating of a fuel channel.