Criminal intelligence report addressed to W. P. Gannaway of the Dallas Police Department. The report, which was submitted by R. W. Westphal and P. M. Parks, regards an interview with Mary Lawrence. Lawrence stated that she saw Lee Harvey Oswald and Jack Ruby meet at the Lucas B & B Cafe after 2 am on November 22, 1963.

Attached is the interpolated uranium burnout values for .570% through 3.000% U-235. Due to increasing interest in higher U-235 values, this document was issued to replace HW-77929 which contained values to 1.519% U-235. The source data was based on the Reactor Cost Studies burnout schedule obtained from Washington -- AEC.

The imminent use of pressurized water-cooled, metallic uranium fuel elements on a large scale in the N-Reactor not only represents an extrapolation from previous Hanford fuel technology, it represents the only production-scale application of this concept in this country or abroad. Provision of a supporting technology is and will continue to be a major activity at Hanford as only a limited amount of directly applicable R&D will be available from other sites. The major benefit to be achieved through an extensive and continuing fuel program is attainment of the full capabilities entailed in the basic fuel concept at much reduced costs. The initial N-Reactor fuel design represents an estimate of what is required for adequate performance. While the design is intentionally conservative, some features may not fully provide the level of performance required to sustain relatively unhampered reactor operation. Production-scale irradiation experience will provide the feed-back required to give direction to a continuing program to correct any excesses or deficiencies in the initial fuel design. Reduction of unwarranted conservatism in the design will lower fuel fabrication costs; correction of deficiencies will lower irradiation costs through increased time operated efficiency. In addition to design optimization, direct savings in fuel fabrication costs …

Investigations of possible operational limitations of the ALSi bonded e elements and the need for a replacement or alternate fuel fabrication process have been lender way for several years. Recently, one fuel manufacturing process, hot-die-sized diffusion bonding, has been intensely examined and irradiation testing of this new fuel type began in 1963. This production test is one additional step in evaluating the new fuel fabrication process. The objective of this test is to evaluate the effect end-bonding techniques on the irradiation behavior of hot-die-sized diffusion bonded fuel elements. Three types of fuel elements will be used in this test: (1) hot-die-sized diffusion bonded elements which have been end bonded by the resistance heating method; (2) hot-die-sized diffusion bonded elements which have been end bonded by the induction heating method; and (3) standard production AlSi bonded fuel elements. Twelve columns of fuel containing the three element types alternated throughout the columns will be irradiated in the test. Twelve smooth-bore Zircaloy-2 process tubes in the C Reactor will be utilized for this test; consequently, all fuel elements of this test are self-support models. These fuel columns will be irradiated to average exposures of 800 Mwd/t. Examination and measurement of the elements of …

At the request of NRD (NPR Physics Subsection) a study was initiated to determine the physics characteristics of various plutonium-uranium composites as fuel for the NPR. From this study, the PuO{sub 2}-UO{sub 2} fuel system was selected to receive major attention. The effect of adding a burnable poison, B-10, in intimate contact with the mixed oxide fuel was also is to be considered. The present report summarizes the results of these investigations.

Research and development activities included: spike loading of N-Reactor, Flex 3 program development, co-product hazards studies, and thorium fuel studies.

This document is a monthly report on the operation and engineering work associated with the N-Reactor department. It summarizes the operating history of the reactor during the month of Nov 1964. Included are brief descriptions of ongoing and planned work on testing and development of reactor components, fuel elements, safety tests, cooling systems, fuel production, materials testing, and other projects to provide safe and efficient plant operation.

This report describes laboratory-scale studies made to evaluate the feasibility of using uranium(IV)-hydrazine instead of ferrous sulfamate-sulfamic acid as the partitioning agent in the Hanford Atomic Products Operation Purex separations plant. Iron(II), in combination with sulfamic acid as a holding reductant, has been used very successfully for many years as a plutonium reductant in solvent extraction separation processes for irradiated nuclear fuels. However, the iron remains in the radioactive wastes produced in the separation processes. Its presence in these wastes accelerates corrosion of waste concentration equipment, complicates treatment of the waste for by-product recovery and adds to the total quantity of contaminated salts which must be stored. Much effort has gone into searches for plutonium reductans which would not have the disadvantages associated with iron(II). Uranium(IV) has been proposed since, having acted as reductant, it is extracted along with uranium from the irradiated fuels and does in solvent extraction separations processes have involved mixer-settler contactors. The HAPO Purex plant has pulsed-column contactors. This difference, plus certain problems due to the particular flowsheet conditions used, required that additional laboratory studies be made to determine the feasibility of using uranium(IV) in the HAPO Purex plant.

Introduction: The Mobile Gamma Irradiator is a semi-production demonstration unit capable of being transported from one area to another for use in irradiating fruit and farm produce, with the capability of furnishing dosages in the pasteurization range.

This investigation was undertaken to determine the approximate impact characteristics of delta stabilized plutonium 1% gallium alloy, both at room temperature and at 325{degree}C. The testing done was non-standard, primarily because some approximate information was desired for another purpose and immediately available materials were used. While the results reported herein should be considered preliminary, they clearly depict the ductile impact behavior of this alloy.

When final system acceptance tests were performed on Project CGI-844, 100-K Coolant Backup System, it became evident that less than the design flow of 32,000 gpm was being achieved. Since the tests indicated that only 28,200 gpm was being supplied to KW and 26,700 gpm was being supplied to KE, it was obvious that the cross-tie lines had higher friction losses than anticipated. A study was therefore made to determine the effects of crosstie line cleaning to remove corrosion nodules. Calculations showed that with the lines cleaned, flows of 31,800 gpm to KW and 30,800 gpm to KE would result. The lines were subsequently cleaned, and on February 16, 1963, flow tests were conducted. The resultant flaws were 31,500 to KW and 30,900 to KE. On November 1, 1963, flaw tests were again conducted to determine if corrosion buildup in the cross-tie line was affecting the flow capacity of the system. While the flow data has not been issued by the test engineers, initial indications are that the present capacity of the system is approximately 29,500 gpm to KE and 30,000 gpm to KW, or a loss of approximately 1500 gpm on each side. The purpose of this document is …

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This report provides a discussion of the co-product production of weapons grade plutonium and tritium in the N-Reactor.

Objective of the KE reactor test is to determine the hydriding corrosion rate of Zircaloy-2 at high temperature in a damp helium atmosphere, and to determine the outgassing rate of tritium generated in a lithium-aluminum alloy. The irradiation capsule design, operating conditions, etc. are discussed.

In an earlier report, the problems associated with the prediction of the critical discharge rates of steam-water mixtures were discussed in detail. That report also gave the results of an experimental investigation of critical flows from long flow passages performed at low pressures. That investigation was later extended to consider short tubes, nozzles, and orifice flow configurations. The present work was undertaken to experimentally investigate the critical discharge phenomenon at higher pressures and to use the data gathered to critically evaluate the various proposed critical flow models. The purpose of this interim report is to describe in detail the apparatus assembled for this investigation and to report preliminary findings regarding the critical discharge rates from a 0.5-inch diameter flow passage.

Quarterly report regarding the activities of the Saxton Plutonium Program, the purpose of which is gathering information and analyzing performance through design, fabrication, and operation of plutonium enriched fuel in a pressurized water reactor system in the Saxton Reactor.

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Declassified 27 Nov 1973. Information is presented concerning Tory-2C operations and testing; reactor kinetics; control system and instrumentation; and physical properties of ceramic materials. (DCC)

This report is to evaluate a silver plated aluminum Bar-X seal's effectiveness in extending the range of ability to seal such a damaged plug.

On the NRX-A reactors the instrument ports are sealed by a Bar-X seal. Damage to the ports, which are part of the pressure vessel, required machining on the pressure vessel. A new design, using a Marman flanged piece to house the Bar-X seal port separate from the pressure vessel, was proposed. The test program was to determine the torque required to seal and the ability of each seal to hold during thermal shock.

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