We have successfully conducted a series of experiments involving scattering of high energy pions and protons from a target containing polarized protons. Results of some of these experiments were reported at this conference, and in the literature. Proton polarizations as high as 65% have been measured; the average polarization during sustained data-taking has been typically 45%.

A research and development project for study of the effects of reactor effluent on Columbia River water quality is being sponsored by the AEC Division of Production. Work was started in October 1962, the first effort being aimed at furnishing an immediate answer to a security question: how closely can Hanford production be estimated by measurement of river temperatures? An interim progress report gave the results of the preliminary investigation. During the calendar year 1963, the study was expanded to meet broader program objectives. This document is a progress report for the year, covering the temperature and effluent distribution phases of the program. Progress on the chemical characteristics phase of the program is documented separately. Figure 1 shows the section of the river under study.

Salt wastes from the Redox solvent extraction process have been routed to the 241-SX tank farm for storage since May 21, 1954. Tanks in this farm contain wastes from three types of irradiated uranium processing: (1) low (approximately 250 MWD/T) and (2) high (approximately 600 MWD/T) natural uranium; and (3) E-metal. It is estimated that approximately 1.6 Kgs of Am-Cm, 17 Kgs of Np-237, 20 Kgs of Pu and 8 tons of uranium are present in the sludges from wastes generated through 4-30-64. A total of 12.1 Kgs of the 29.1 Kgs of Np-237 estimated has been recovered or is present in the Redox Np-237 accumulation cycle as of 4-30-64. Because of the potential for unmeasured losses (such as Pu or U precipitates) the total plutonium and uranium may exceed the quantities stated herein. A breakdown of the estimates by tanks is provided.

Four years of intermittent development has culminated in the successful test drilling on-reactor, of one step plug hole. Since 1960, several different drive units and many diamond, carbide, and tool steel cutting heads were tried unsuccessfully in attempts to bore a stepped hole in a mockup of B, D, P side shielding. Success was finally achieved in 1963 using a standard horizontal boring mill and tool steel cutters. With slight modifications, this same equipment was successfully used in an on-reactor test drilling at F Reactor in December of that year. The on-reactor test revealed the need for improvements in the equipment. Chip flow was inadequate and appeared to be caused by poor air flow in the core receiver. Cast iron cutting technology used on the mockup did not work on-reactor and had to be revised on the spot. The graphite did not break up into chunks as desired and had to be manually removed. None of the steel cores were more radioactive than 50 mr/hour at approximately 2 inches, but the cast iron core read 300 mr/hour at 10 feet. It had to be handled very quickly. The dose rate for handling the graphite was 400 mr/hour. It, too, was …

Information resulting from recent graphite investigations in the K Reactors were presented to the Council. Separations as much as three to four inches were reported. Contraction of the graphite within the fueled regions of these reactors is now about one per cent in the direction transverse to the extrusion axis of the graphite and about two per cent in the parallel direction. Saturation is not expected until shrinkage of at least two to three per cent is observed in the transverse direction. The contraction is resulting in two principal operating problems; control rod entry and loss of 3{times} balls in the separations following a ball drop. The principal purpose of the review concerned only the ball system and particularly the philosophy governing the system`s use. Following a historical review of the ball system philosophy, our present views were developed by the consideration of six questions and answers. Our points extracted from these questions were (1) the IPD reactors require a backup safety control system (2) the backup system should be separate from the primary safety control system, (3) the backup system should be activated automatically upon accidents capable off disabling the safety rod system, (4) the system actuation should be …

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

EG&G submits the following estimated cost at completion of the ETS-1 I&C System as represented by the approved, Reduced Design Package, the SNPO-C letter TSB:AL dated 30 January 1964 and the Approved Program Plan. A summary of this cost estimate and of previous cost estimates, is given. The total cost estimate is $5, 284, 000. While some items have individual costs which differ from those shown in SNPO-C letter of 30 January 1964, the total cost is consistent with that letter. Specifically, there is a greater proportion of equipment costs in this estimate than in the funding plan contained in the SNPO-C letter due to new equipment estimates from industry. We discuss the basis for the present cost estimate, and summarize options and leasing plans which will be exercised by the Government. We also discuss Facility Modification design and Miscellaneous Crafts Work, respectively. EG&G has only design responsibility for these latter two items. The cost estimates are best estimates, neither worst cases nor optimistic cases, and do not allow for contingencies. While contingencies have not been included in previous Cost at Completion Estimates, we believe that in the spirit of the second paragraph of SNPO-C letter dated 30 January 1964, …

The Redox acid precycle flowsheet which has been in use since November of 1959 vas adopted to provide neptunium recovery. Although neptunium recovery on this flowsheet has been satisfactory, several problems have been encountered. Among these are: marginal uranium decontamination, increased hexone degradation and increased corrosion of precycle piping and equipment. To circumvent these problems an acid-deficient precycle flowsheet was proposed. In August, 1964, following replacement of the 1A Column and the D-14 backcycle concentrator, the Redox plant was started up on this flowsheet. Neptunium, which left the 2A Column in the waste stream was recovered by processing the waste stream through the 1S Column on an acid flowsheet. The neptunium, extracted in the 1S Column was stripped in the 10 Column and routed to the D-14 backcycle concentrator in the 10W stream. Using this flow pattern it was planned to accumulate neptunium in the D-14 and F-8 vessels-and the 1A, 1S and 10 Columns. Inadequate stripping of neptunium in the 10 Column with resultant neptunium losses to the waste stream necessitated abandoning this flow pattern and the neptunium stream from the 1S Column has been routed to the 2E Column with the 1A Column product stream. Neptunium is presently …

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The purpose of this test was to evaluate the behavior of natural lithium-bearing materials and to provide samples for production and extraction information. Testing of the target elements was done in conjunction with a test evaluating the behavior of NIEl fuel elements. Conditions for operation during the test were similar to, or more severe, than conditions expected in N Reactor.

The attached tables are an extension of the data that will be used in the forthcoming Production Studies. It considers thorium load substitution for the natural uranium cases previously reported. The power level, operating efficiency, exposure, conversion ratios, tonnage, and production are given for the natural U loading, blanket loading, and full conversion. The small reactors, the C reactor, and the K reactor are considered.

At this present time, fuel elements of two basic lengths are used in the IPD reactors --a ``six-inch`` enriched uranium element (0.947 w/o U-235) and an ``eight-inch`` natural uranium element. Although fuel element diameters vary fro model to model, the length of the uranium cores is reasonably constant. An evaluation of fuel element lengths has bee made and the economic incentive for producing and irradiating longer fuel elements is fairly large. This test is a first step in evaluating the irradiation behavior of a new fuel model. The objective of the production test detailed in this report is to evaluate the irradiation behavior of ``ten-inch`` enriched (0.947 w/o U-235) uranium fuel elements.

A series of arc-jet tests were performed for the purpose of investigating the behavior of SNAP 2/10A fuel in a simulated reentry environment. The data from the tests (motion pictures and transient temperature distributions) were analyzed and compared with computations which used an energy balance to describe the heating of the fuel to its melting temperature. The dissociation of hydrogen was assumed to occur as an isothermal, endothermic phase change. This concept neglects the kinetics of diffusion. (A computational model incorporating these kinetics was being developed concurrently with the test.) A temperature of 2000/sup 0/F was assumed for the dissociation phase change and an alternate temperature of 1800/sup 0/F for comparative purposes.

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Water for fire and sanitary use in the older reactor areas is provided by pumps located in the 183 building of each particular area. Each area has two electrically driven pumps for normal use and two steam driven pumps in case of power failure. Distribution within an area is through an underground main loop with laterals to buildings and fire hydrants. When the DR, F and H Reactors are deactivated, fire and sanitary water in each area is still necessary for protection of the area and for certain peripheral facilities which can not be moved immediately to other locations. In D area, deactivation of the DR Reactor will not effect the fire and sanitary water system but in F and H areas, deactivation of the reactors will cut off the source of fire and sanitary water unless 181 river pumps, 182 filter plants and the 184 power houses are kept in operation. The purpose of these design criteria is to describe the scope of work and operational and technical requirements for conversion of the export water line between D and F areas to a fire and sanitary main to supply water from D area to F and H areas. 2 …