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From introduction: "Summary report of the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company."

From introduction: "Presents the detailed description of a 1000 MW Electric Power Plant employing one or two Boiling Water Reactors as the steam source."

From introduction: "Contains the appendices and a complete set of drawings related to the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company for the United States Atomic Energy Commission."

Isotopic assays have been made on the Cr in samples from 14 different chrominiferous minerals from different geographic and meteoritic sources. The results of the assays indicate that it is not possible to state unequivocally that variations in isotopic composition have been observed. (auth)

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The potential of shipping a bucket of metal that has not had sufficient decay time is a common problem to all Reactor Processing Operations. The only method at present to assure shipping adequately decayed metal is by procedural control. This study was performed to determine the feasibility of establishing decay time for irradiated metal in buckets by radiation activity measurements.

One of the characteristics of the safety control system of the Hanford reactors which is of primary concern from the standpoint of reactor safety is the response time. The purpose of this study is to determine whether the safety control system is sufficiently fast-acting to prevent an intolerably large power excursion from developing. An intolerably large power excursion is one that would cause a melting of the uranium while the reactor power is above the previous operating level. The purpose of this report is to show the techniques used to analyze the speed of control of the Hanford reactors. This report will help to standardize the methods and reduce to a routine the solution of this type of problem. The analog circuits will be presented and explained. A report has been written previously on the speed of control. The changes in the problem have necessitated the redesigning of the analog circuit and also the issuing of this report. This report compiles and organizes the analog work done on the speed of control to date.

A study was made to determine the structural and functional adequacy of the EGCR graphite core design. Maximum stress and distortion of the core and the probable operating time before graphite cracking were determined. The major cause of stress is non-uniform fast-neutron flux, which causes non-uniform shrinkage of graphite components. The critical stress was found to be tensile. The criterion for determining the time of cracking of graphite columns appears to be the point at which the creep strain results in rupture. Column life before cracking may also be related to the maximum allowable stress at rupture. It was concluded that non-uniform shrinkage is the major cause of distortion of graphite components of the core. In general the amount by which a column tends to bow during 20-year core life exceeds the available free space. The combined internally generated and bowing stresses result in an average period before cracking of all core columns of 14 years, with a minimum period of 6 years. (M.C.G.)

Measurements of thorium resonance-capture activations relative to those of Au/sup 197/, which were made in the Peach Bottom (HTGR) critical assembly, are analyzed and compared with thorium resonance activations calculated from tabulated values of the resonance parameters. The new vanadium-subtraction method that was applied is shown to measure thorium resonance captures relative to those of Au/sup 197/ to within several per cent in this geometry. The measurements show that the lowest several resonances of thorium capture less strongly than would be predicted by the listed parameters. This reduced capture is such that the thin-limit, epicadmium 1/E-spectrum resonance integral is about 84 b instead of the 96 b predicted by listed parameters; the thorium resonance captures for the critical assembly are about 6% less. The approximations currently used in calculating resonance captures in annular lattices probably underestimate captures by several per cent. The 300 deg K resonance-capture rate of ThO/sub 2/ in the annulus is indicated to be several per cent larger than that of metallic thorium, presumably because of a solid-state effect. 46 references. (auth)

Figures are included which gives the necessary information on C Reactor for analysis of stress buildup in the graphite tube blocks.

Methods for dissolving unirradiated uranium-molybdenum alloy reactcr fuels in nitric acid, nitric acid--ferric nitrate, and nitric acid-- phosphoric acid solutions were studied on a laboratory scale. Flowsheets based on the results propose dissolution of alloys containing 3% molybdenum in boiling 6 M HNO/ sub 3/ to yield stsble solutions that are 0.6 M in uranium and 3 to 4 M in nitric acid. The uranium can then be easily decontaminated and recovered in a conventional Purex-type tributyl phosphate solvent extraction process. Alloys containing 10% molybdenum would be dissolved in boiling 11 M HNO/sub 3/, allowing molybdic oxide to precipitate. The molybdic oxide, which carries 5-10% of the uranium, is removed by centrifugation and the acidity of the supernatant solution adjusted tc allow recovery of the uranium by Purex-type solvent extraction procedures. The uranium carried by the molybdic oxide is recovered after the MoO/ sub 3/ is dissolved in warm 5 M NaOH. Less than 0.1% of the uranium is solubilized during the caustic dissolution. Alternative methods investigated involve dissolution in nitric acid containing 0.5 to 1 M ferric nitrate to complex the molybdenum. These techniques lead to undesirably large volumes of high-level solvent extraction waste solutions. Phosphate ion is also …

The status of aqueous processing methods for thorium fuels is summarized, with principal emphasis on the stainless steel-clad ThO/sub 2/UO/sub 2/ type. Data were obtained principally from laboratory-scale experiments with fully irradiated fuel samples and engineering-scale tests with unirradiated fuel. Stainless steel cladding was easily dissolved with 4 to 6M H/sub 2/SO/sub 4/ (Sulfex process) or 5M HNO/sub 3/-2M HCl (Darex process) in LCNA (Nionel type) or titanium equipment, respectively, in semicontinuous or batch equipment. Uranium losses to the decladding solutions were approximates 0.3% and 3 to 5% for the Sulfex and Darex processes, respectively, with fuel irradiated to approximates 20,000 Mwd/ton of core. The uranium was readily recovered from the Darex decladding solution in the acid Thorex extraction process. The ThO/sub 2/UO/sub 2/ core was dissolved in 13M HNO/sub 3/ -0.04M NaF-0.1M Al(NO/sub 3/)sub 3/. Uranium and thorium can be recovered from graphite-base fuels by disintegration and leaching with 90% HNO/sub 3/, grinding and leaching with 70% HNO3, or combustion followed by dissolution in fluoridecatlyzed nitric acid. Uranium and thorium were recovered from nitric acid solutions and separated from fission products by extraction with 30% tributyl phosphate in Amsco in the acid Thorex process. The use of an acid …

Research and development on metallurglcal aspects of pressurized-water systems is summarized. A survey was made of the methods of determining fuel burnup. The mechanisms and kinetics of the loss of boron during heating at 1135 deg C in various dynamic environments were determined. A model was developed to quantitatively characterize the UO/sup 2/ dispersion microstructure of roll-clad fuel plates relative to an ideal'' dispersion. In order to avoid the loss of boron from UO//sub 2/- stainless steel dispersion fuel plates during fabrication, studies were carried out on a refractory glass containing 4 wt.% B/sub 2/O/sub 3/. By using lowsilicon elemental powder, the undesirable reaction between Eu/sub 2/O/ sub 3/ and Si was eliminated; and 13 full-size SM-1 absorbers were fabricated. Work was continued on the borongradient neutron absorber concept. A design was studied for preparing a composite control rod having an upper section made of a boron-gradient dispersion and the lower tip made of Eu/sub 2/O/sub 3/ and stainless steel. Two fuel elements were examined after significant exposure in SM-1. The examination of the miniature boron-iron samples in the final phase of the MTR irradlation test was performed. Twelve miniature test specimens containing 20, 30, or 40 wt % Eu/sub …

This report addresses a survey of background and legal problems.

The effect of extracted zirconium and nitric acid upon the chemical stability of tributyl phosphate was investigated using gas-liquid chromatography. Tributyl phosphate was degraded approximately 1000 times faster by the tributyl phosphate-zirconium reaction than by the tributyl phosphatenitric acid reaction. Normal butyl nitrate was the major volatile product for both systems studied; with extracted zirconium a solid complex corresponding to the formula Zr(NO/sub 3/ )/sub 2/(DBP)/sub 2/ was also obtained. (auth)

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Budget figures are given for N-reactor department, irradiation processing department, chemical processing department, and Hanford laboratories.

Discussion of laboratory tests to reduce uranium corrosion.

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This report discusses how the co-sponsorship of bills and resolutions work in the House of Representatives.

At your request the Plant pressure monitor scram history for the period covering January 1, 1960, through June 31, 1963, was reviewed to predict the production saving that would result if a reliable pressure monitor dual-trip feature were added to the six old reactors.

To permit the irradiation of one dummy fuel element assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

This report addresses the compact thermoelectric converter from April 1 to June 30, 1966.