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From introduction: "The primary objective of the present study was to develop a conceptual design of a large sodium cooled fast breeder reactor of a nominal electrical rating of 1000 Mw operating on the uranium-plutonium cycle."

Report containing plans for a desalination pilot plant for the purposes of evaluating the feasibility of reverse osmosis for the desalination of sea and brackish water with the capacity of 1,000 gallons of potable water produced per day.

Work was continued on the development of a process for the gasification of bituminous coal with heat from a nuclear reactor. Major objectives of the project are: (1) develop and test components for a gas-cooled reactor system that can heat gases to 2000/sup 0/ to 2500/sup 0/F; (2) investigate coal gasification methods compatible with this system; and (3) develop an exchanger that can utilize high-temperature heat to gasify coal or heat chemical process streams. During the past year, work was concentrated on the design, construction and installation of a fluidized-bed gasifier in the high-temperature heat system. Design of the fluidized-bed gasifier was based on information obtained from laboratory and pilot-scale gasifiers. The gasification chamber, a length of 6-inch, schedule 80 stainless steel pipe, is fitted with ten 1/2-inch Hastelloy-X tubes running lengthwise. Hot helium flowing through the 1/2-inch tubes provides heat for the gasification of the steam-fluidized char in the 6-inch pipe. Helium leaving the gasifier generates the superheated steam required for gasification. Fabrication and installation of the gasifier was completed and shakedown runs started. In the gasifier, the heat-exchanger tubes will be heated to 1900/sup 0/F in a mixture of the corrosive gases, including H/sub 2/, CO, CO/sub 2/, …

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.

Cerium sulfate purified by D2EHPA in Semiworks can be precipitated by adjusting pH to between 1 and 2 in tank 6 with 50% caustic. The solution can then be transferred through tank 1 to tank 67, where sodium bisulfate is added to make the solution 0.5M sulfate. A stoichiometric amount (mole for mole) of 50% caustic is added to just neutralize the sodium bisulfate. The precipitate is digested one hour at 60 C, then filtered.

This document details activities of the irradiation processing department during the month of December, 1963. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; and Financial Operation.

The monthly report for the Hanford Laboratories Operation, December 1963. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics, and programming operations are discussed.

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.

Studies initiated at Hanford in 1961 for the purpose of developing an alternate cladding process for I&D fuel elements for the eight existing production reactors indicated that the hot die sizing diffusion bonding process offered the greatest incentive. Hot die sizing was the most attractive with respect to improved fuel quality and potential reduction in fuel element unit cost when compared to the existing AlSi brazing process. Initial development work consisted of determining optimum process parameters for producing good diffusion bonds on the inner and outer lateral surfaces during sizing. This report summarizes the results of a series of end bonding variables tests designed for producing good diffusion bonds on the ends of hot die sized fuel elements.

Here is a more complete description of the capsule test we discussed in our initial contact in December. I have included as much detailed description of the test as has been decided on to date. Also, there are some factors which, from previous similar test, I have found to be pertinent to the successful charging and irradiation performance of the capsules. There is one critical point that needs to be settled as soon as possible. In order to finish machining the outer diameter of the capsules, the expected approximate specific power generation and enrichment in the surrounding process tubes must be known. In addition, there are several other factors which effect the location of the test in the reactors which Bob Marshall and I would like to go over with one of hour engineers as soon as possible.

The steadily increasing graphite stack distortion in the K Reactors has caused serious operating problems with the existing horizontal control rods. To compound the seriousness of the problems, the high level of reactor operation today and the anticipated higher level of operation in the future demands a reliable control rod system. A flexible control rod has been designed by Reactor Design, IPD, to facilitate reliable operation of the HCR system in spite of channel bowing arising from graphite stack distortion. This flexible control rod design is radically different from the existing K Reactor control rods and in fact, is more closely aligned to the control rods now in use at the older Hanford Reactors. The major difference of the new rod is the elimination of intimate contact between the poison-containing section of the rod and the cooling water. Such a change in design as described above could result in significant changes in the operating temperatures of the rod proper. This study was undertaken to provide a calculated indication of the temperature changes and the relative magnitude of such changes relative to reactor power levels, graphite temperatures, coolant temperatures, etc. In addition to this basic information, the scope of the study …

This invention report suggests a method for drastically increasing the amount of Pu-241 in isotopic mixtures of plutonium. It is felt that with process experience, as much as 70 percent or more of the potential Pu-241 atoms can be concentrated in the fuel at one time. Such a concentration step would double the amount of production obtainable from Pu-241 by nuclear decay. The process to concentrate the Pu-241 consists of two basic steps: 1. Irradiate plutonium consisting largely of Pu-239 isotope until the Pu-239 has largely been converted to Pu-240 and 241 by thermal neutron fission and absorption events. Most of the fuel value has been taken advantage of at this point. The isotopic mixture will be largely Pu-240 but contain smaller percentages of 241 and 239. The irradiation is terminated at this point and the products are separated. 2. The depleted plutonium isotopes from the first irradiation are refabricated for a second irradiation with a thermal neutron absorber surrounding the depleted plutonium isotope. This element is again irradiated, preferably in an epithermal neutron flux with a peak energy slightly above 1 ev. The Pu-240 has a huge resonant cross section at the 1 ev level and will be most …

This report details activities in present reactors and N-Reactor fuel development.

Early in December 1963 eighteen natural uranium columns were discharged from C Reactor. Fifteen (15) of these columns contained alternately charged HDS (test) and AlSi (control) fuel elements in the downstream half (positions 1--16); two columns were charged full length with HDS material; and one column was charged full length with AlSi material. All canned pieces were nominally C5NS dimensions except that the HDS pieces were slightly longer than the standard AlSi pieces. Uranium fabrication history, through heat treatment, was controlled ad equivalent for both test (HDS) and control (AlSi) material. For these eighteen columns average exposure was {approximately}960 Mwd/ton, average tube power was {approximately}1125 kw, and average tube outlet temperature was {approximately}100 C. Two striped charges were discharged in September 1963 @ 370 Mwd/ton. This report presents results of the post-irradiation measurements that have been completed and analyzed as of this date. A second set of measurements for a portion of the material is being programmed.

Normally, in the flow of a flashing liquid through a pipe, it is assumed that the location of the choking phenomenon, if it exists, is at the exist of the pipe. Experiments have indicated that thee are circumstances under which choking can occur at the entrance instead. These experiments were performed to study the flashing of initially-compressed water through short lengths of pipe with length/diameter ratios of up to 20. Entrance configurations considered included sharp and rounded transitions and conical passages of 20- and 45-degree tapers. Upstream pressure ranged from 115 to 1,800 psia and in all instances, the upstream pressures were much greater than the saturation values. In most cases, choking of the flow was incurred when the liquid entering the pipe was required to depressurize to the saturation pressure (or below) when accelerating through the reduction at the entrance. Axial pressure surveys demonstrated that the pressure at the entrance of the pipe during choked conditions was the saturation pressure. It remained at the saturation value regardless of upstream pressure and the pressure distribution along the remainder of the pipe.

The objective of the production test described in this report is to authorize the irradiation of aluminum-clad fuel elements with U{sup 235}-Al cores for corrosion testing in the 1706-KE Facility; the irradiation of thermocouple elements to measure the temperature of the aluminum cladding, installation of modified rear nozzles on the tubes in which these tests will be conducted., and the use of two regular KE tubes as controls. In order to measure corrosion at higher heat flux and surface temperature than are ordinarily available, corrosion measurement will be made using enriched aluminum-clad fuel elements with ``C`` and ``J`` (U{sup 235} in Al) cores. One fuel element per charge will be provided with thermocouples inserted in the cladding to measure its temperature. These tests will be conducted in the ribless zirconium single-pass tubes in KE Reactor which can be supplied with water from 1706-KE. These fuel elements will have special end supports to avoid disturbances with flaw and heat transfer on the lateral surface. Modified rear nozzles, of the type designed for use with K zirconium tubes, will be installed to permit smooth discharge of elements with these supports.

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It had been postulated that the increased film area of a charged overbore tube would be counterbalanced by the lower temperature fuel surface and decreased flux so that the radioisotope content of the effluent would be the same as that from a normal tube. In order to confirm or deny the postulate, Tube 3267-C was equipped with an effluent sample line under the provisions of PT-IP-543-A. Sampling and analytical work began in mid-April and continued through November, 1963. Effluent samples obtained from an overbore tube at C Reactor over a nine-month period were analyzed for P{sup 32}, As{sup 76} Np{sup 239} and Cr{sup 51}. Based on the results obtained for these isotopes, it can be concluded that the radioisotope content of the effluent from an overbore tube will be the same as the radioisotope content of the effluent from a normal size tube operating under similar conditions.

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Abstract: This report gives a general description of the single stage axial compressor blade test facility located at the Oak Ridge Gaseous Diffusion Plant, Oak Ridge, Tennessee.

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 …

Studies of reactor safety considerations have demonstrated the need for automatic safety circuit action based on the measured rate of power increase in the power level range from 10{sup {minus}2} to 10 {sup 0} times equilibrium power level. Budget and Preliminary Engineering Studies were performed. This document provides the design criteria for detailed design of the proposed Linear Power Rate-Of-Rise Instrumentation facilities and is applicable to any of the eight IPD reactors.

This report, for December 1963 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Financial operations; facilities engineering; research; and employee relations. Weapons manufacturing operation; and safety and security.