This report addresses the first general analysis of the circular loop as a transmitting antenna, which appears to be that of Hallen.

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The status of aqueous processing methods for tharium fuels is reviewed. A specially designed 250-ton shear was successfully tested for shearing full-size simulated Consolidated Edison type unirradiated assemblies into 0.25-1.5 in. lengths.

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The design and characteristics of the Janus Reactor Facility for medical and biological research are described in detail. Pertinent information is given about the nuclear and reactor radiation characteristics of the facility. Information is presented on the design and operation of reactor components, auxiliary equipment, instruments, and associated equipment. Maintenance and operating instructions are included for the facility. Cost breakdown is outlined. (N.W.R.)

This report addresses the semiannual progress report for the period ending on February 15, 1963.

Tests were run to determine the burnout conditions for a nonuniformly heated rod in an annular geometry. Two special electrically heated rods were used; the cosine rod, which gave an approximate cosine axial heat flux distribution; and the truncated cosine rod, identical to the cosine, except for one end cut short. The rod to be tested was placed in a circular tube test section to form the annular flow path for the water coolant. Only the rod was heated; the outer surface (tube) was essentially adiabatic. Orientation was vertical, with flow upward. The tests were run at the following conditions: rod OD, C.540 inch; tube ID, 0.875 inch; hydraulic diameter, 0.335 inch; heated length of cosine rod, 108 inches; heated length of truncated cosine, 91 inches; pressure, 1000 psia; flow rate, 0.84 x 10/sup 6/ to 1.40 x 10/sup 6/ lb/hr-ft/sup 2/; and steam quality at exit, 12 to 35%. For each flow and iniet subcooling, the electrical power was increased until burnout was reached, thus establishing a burnout condition. Each nonuniform rod was instrumented with thermocouples in the region of anticipated burnout. The local heat flux and local quality at each position at which a thermocouple indicated a temperature …

A CDC-1604 subroutine package was written to facilitate the plotting of curves and points on linear, logarithmic, and semilogarithmic graphs using the CALCOMP plotter. The subroutines accomplish the necessary computations and prepare a magnetic tape for use by the plotter. (auth)

The conduct of outage work in the discharge areas of the Hanford production reactors has always been complicated by personnel exposure to radiation from the contaminants which accumulate in the effluent water piping in the course of normal operations. These containments represent trace impurities in the cooling water, together with corrosion products, which are irradiated in passing through the reactor and then deposit in the unshielded rear face piping. Radioactive residues also accumulate on the external surfaces of this same complex and congested hardware. The radiation from these depositions increased as power levels and effluent water temperatures were raised, and by 1956 it was apparent that corrective steps would be essential to effective future maintenance work in the discharge areas. It is the purpose of this report to summarize the subsequent development and application of the chemical process for rear face piping decontamination which is being used so effectively in connection with the tube replacement program and other major outrage work.

This report, from the Chemical Processing Department at HAPO for May 1963, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; weapons manufacturing operation; and power and crafts operation.

This document describes the detailed test procedures to be used during the initial phases of the physics test program--the approach to critical and minimum critical size tests. During these initial phases, several physics characteristics of the reactor will be evaluated in addition to determination of the critical loading. Overall flux distributions will be determined in the critical pile and used to evaluate the lattice buckling characteristics. Theoretical calculations normalized to experimental values from exponential and PCTR tests have been used for estimating the experimental results for each phase. The startup tests will be used to verify the preliminary calculations and to define the reactor characteristics within the precision required for safe and efficient operation of the reactor. Detailed and accepted test procedures are required for maximum test efficiency and reactor safety during the physics test program and for sufficient overall scheduling into the N-Reactor operational test program. Hazards` analyses have been performed and are presented for cases which are unique to the physics startup tests or significantly differ from the Design Hazards Review.

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Experimental parameters of ZPR-HI fast reactor assemblies were compared with those calculated through use of a cross-section set derived from that of Yiftah, Okrent, and Moldauer, and initially used for a study of the critical sizes of ZPR-III assemblies. This cross-section set waa also used to study central fission ratios in ZPR-III assemblies. The analysis was extended to prompt neutron lifetimes and the relative reactivity effects of the substitution of materials at the reactor centers. The results of the work were examined in conjunction with those of the critical size and fission ratio studies. (auth)

Verify that the multigroup code P1MG and its associated library yield, for a flat flux problem, the same few group constants as the MUFT fast constant code.

Modifications consisted of redesigning the feedback loop and shielding the toroid to minimize capacitive pickup and of insulating the core from the magnet buss. A system analysis is presented, and system controls and operation are described. Operational properties of the redesigned loop and tests on the insulation between the grounded Faraday shield and the magnet buss and core are discussed. (D.C.W.)

Over the past few years we have developed a system which allows several different liquid hydrogen and liquid deuterium targets to be placed in the electron beam of the Stanford Mark III accelerator and which can be operated from a remote position. The geometry of our scattering chamber requires that the motion (of the various targets into position) must be in the vertical direction, but that the overall height of the system be constant. These requirements are met by using an internal retractable bellows arrangement to raise and lower the target cells. As many as five cells can be used in conjunction with a single reservoir system.

Criteria provided in this report delineate the objective, bases, and functional requirements that shall govern the preparation of detail design of the bauxite-sulfuric acid feed facilities to be installed in the 183-KE and KW Buildings. These facilities will produce the chemical coagulant used in the treatment of Columbia River water in the water plants and thus replace the existing liquid alum feed systems used for this purpose. The objective of this document is to define the operational and technical requirements of the new process and to outline the functional requirements of the proposed facilities for the purpose of detail design. The criteria below define the requirements for a single K Area water plant. Unless otherwise stated they shall apply for both K Area water plants.

This document presents the initial definition of a 2 Mwt, SNAP-50 Reactor Test and is intended and is intended to provide guidance to the Architect/Engineer in evaluating and costing various Nuclear Test Facility design concepts. The information is based on preliminary studies; and, consequently, the configuration and component sizes have not been firmly established.

This document presents the initial definitions of an 8 Mwt, SNAP-50 Reactor Test and is intended to provide guidance to the Architect/Engineering in evaluating and costing various Nuclear Test Facility design concepts. The information is based on preliminary studies; and, consequently, the configuration and component sizes have not been firmly established.

The component test program has been designed to permit a realistic evaluation of critical components and, where necessary, to point the way to design or operational changes which will provide a high level of assurance of a successful power test.

A study was made on the Ti--Al-Nb system for high temperature alloys with useful strengths above 1200 deg F. Composition variations included: niobium from 17.5 to 35%, aluminum from 12.5 to 17.5%, and the addition of smaller amounts of zirconium, hafnium, tin, carbon, or boron. Short-time tensile properties were determined for one strain rate (0.0029 in./in./min.) at 1600 deg F. Stressrupture properties were obtained for a variety of temperatures and stresses. The Larson-Miller parameter was used to present these stress-rupture data so that the rupture life at various temperatures could be presented on a single plot for each alloy. (auth)

Work was done on the adaptation of an electrical probe, developed by Neal for the measurement of local void fractions in mercury-nitrogen flow, to air- water flow. The adaptation is more difficult because of wetting of the probe by water. Various probe shapes and filming agents were employed without great success. Some improvement was obtained by means of a separate triggering circuit, but the calculated void fraction was still somewhat low compared with that measured by the gamma-ray-attentuation technique. Presumably, this is due to a finite response time of the probe, associated with the wetting of the tip; hence, further development of this technique is necessary. (auth)

The tensile and yield strengths of both cast and wrought uranium discs were substantially increased by explosively impacting them at room teraperature and at 375 deg F. However, the room-temperature impacting caused gross damage in the cast material and slight internal damage in the wrought material at the highest impacting pressures. Impacting at 375 deg F, which is just above the brittie-ductile transition temperature for uranium, was the most effective method for increasing the strengths with no damage to either the cast or wrought material. This impacted material retained some of its increased strengih after a lowtemperature (425 deg C) vacuum anneal that greatly increased the elongation. A salt anneal caused a partial recrystallization in the impacted cast uranium. (auth)

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