Report discussing the mechanics of X-Ray and COMBO computer programs.

The recently completed 25-inch hydrogen bubble chamber combines excellent picture quality with a fast operating cycle. The chamber has a unique optical system and is designed to take several pictures each Bevatron pulse, in conjunction with the Bevatron rapid beam ejection system. At present the chamber operates twice per Bevatron pulse. The general features of the chamber constructiosn are shown in Figs. 1 and 2. The chamber is ten inches deep at the narrowest point. The magnet is of conventional water cooled design and in present operation produces a vertical field of 18.5 kgauss. With suitable generators it has produced a field of 22.8 kgauss. The most important new features are discussed below.

Program 3 of Operation BREN consisted in a study of the performance of biological shields composed of various arrangements of steel, water, and borated polyethylene in spherical, cubical, and parallelepipedal geometries. Studies were also made of the radiation levels within concrete-lined and -capped pits, 4 ft in dia and 8 ft deep. The principal radiation source was the ORNL Health Physics Research Reactor, an unshielded, unreflected, 90 wt % U-10 wt % Mo, cylindrical bare-metal assembly. It was supported at heights ranging from 27 to 1500 ft by a hoist car riding outside of a 1527-ft-high steel tower. A 1200-curie Co/sup 60/ gamma-ray source, similarly supported, was substituted for the reactor for some measurements. The work reported includes measurements of fast-neutron and gamma-ray dose rates and thermal-neutron fluxes in air and within the various shields, using the reactor source, measurements of gamma-ray dose rates in air from the Co/sup 60/ gamma-ray source, and measurements of gamma-ray pulse-height spectra within the pits, also employing the Co/sup 60/ source. 8 refs., 33 figs., 29 tabs.

This document details activities of the Irradiation Processing Department during the month of June, 1964. 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; Employee Relations Operation; and Financial Operation.

Efforts have been made to experimentally evaluate and to calculate the effect of reactor environment on splines. This report follows some of these efforts and relates them to operational experience. The use of splines in the Hanford Reactors has improved the efficiency and ease of operating the reactors. With the widespread use of splines, it is important to know the effect of reactor exposure to safely and efficiently use the splines in reactor operation.

Neptunium-237 production has been-calculated according to the equations documented in Reference 1 for use in 1964-AEC study. cases. The assumptions made in the forecasts are as follows: Only the 947 stream is recycled. The blend enrichment for the recycle maternal is 1.2 per cent for F-Y 64 and 65. For FY 66 -73, a blend enrichment of 1.4 per cent is assumed which is better value according to the optimization model of C. W. Showalter. The 947 stream recycle time is two years. In cases A and B where the 947 stream volume is materially increased, virgin 947 material is introduced to make up the material balance. The U-236 content of the blend material- is 66 ppm{sup 2}. The U-236 input level in the 947 stream for the input of FY 64 is 200 ppm{sup 1}. Separations recovery is 85 per cent for FY 64--65 and 90 per cent thereafter. The material throughput data are given in Table I. These values are consistent with those being assumed for the AEC study cases. The Kgs of Np-237 by fiscal year are given in Table II. It must be remembered that these forecasts are based on a particular set of idealized assumptions, …

Tube life estimates for the base case and the five study cases are given for 6.6 and 7.0 pH. The fuel sizes and operating conditions for each case are summarized. Information on the mixer position that gives maximum tube life for each case is shown. A tube life distribution was calculated for the base case using the outlet temperature distribution for 1197 central zone tubes at B reactor. Average tube life for the 1197 tubes war estimated to be 3.0 years (100 percent TOE) for operation at 6.6 pH, and the outlet temperature corresponding to average tube life vas l07{degree}C. Similar numbers could be calculated for the study cases if estimates of the outlet temperature distributions were available.

This report, for June 1964 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.

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Irradiation of thorium for the production of ``clean`` U-233 at HAPO began in CY 1963. Several small-scale tests, involving from gram-size quantities up to a few elements, were successfully conducted; the results from these tests, coupled with theoretical predictions, indicated the desirability of further investigation. The objectives of this test are to (1) authorize the irradiation of about two tons of thorium oxide target elements in central zone process tubes of KE Reactor, (2) determine the reactivity effects of an enriched uranium-thorium oxide (E-Q) loading in the K Reactors, (3) demonstrate the feasibility of using the K Reactors in production of ``clean`` U-233, (4) provide sufficient irradiated thoria material to permit demonstration of the chemical processing capabilities, and (5) obtain data from an extended irradiation of one column of thoria elements.

The General Electric Company, as contractor to the AEC at Hanford, is proposing to modify the fuel loading of one or more of the IPD production reactors for the purpose of producing ``clean`` U-232 as a coproduct with plutonium. The isotope, U-233, with low (3-5 ppM) U-232 content, is expected to become a material with important nuclear applications. The IPD reactors are well suited to produce clean U-233. A proposal to produce 200 kg of U-233 for critical experiments for the AEC`s seed-and-blanket reactor program have been made in response to recent AEC inquiries. The production of such large quantities of U-233, on the schedule requested, would require nearly full utilization of the IPD reactors for a period of about half a year. This report, therefore, is intended to evaluate the nuclear safety of the IPD reactors whose entire loadings have been altered to coproduce U-233 and Pu-239. The coproduct loading will involve irradiating thorium oxide target elements. Excess reactivity to support the target material will be furnished by slightly enriched uranium, charged in tubes separate from the thoria. The expected fuel-to-target ratio is about 5.5 in the reactor cores and two in the reactor fringes. A full technical hazards …

This report provides the Redox Operation material balance and production for fiscal year 1965.

The N-Reactor Hazard Summary Report examines the hazard from operating the N-Reactor with a uniform fuel loading enriched to 0.947% U{sup 235}. Incentives have been developed for reactor testing of a block of 49 tubes loaded with co-producer elements, i.e. elements capable of producing both weapons grade plutonium and tritium. The element utilizes an outer fuel tube enriched to 1.25% U{sup 235} with an inner target lithium-aluminum rod. Criteria have been developed to guide the evaluation of safety aspects of such tests. It is the purpose of this document to review the hazards associated with the proposed test and to set forth special precautions which will be necessary to maintain a high level of safety.

This document is a summary of the new Reactor Operations Daily Report, which is to be implemented August 3, 1964. The report explains how the form is to be filled out, and what information is to be entered into the different fields on the form.

The N-Reactor was shutdown on July 7, 1964, due to fuel rupture indications on tube 3361. The rupture has been confirmed as a failure in the downstream inner fuel element in tube 3361 caused by a pin hole in the weld and braze metal. This document presents a historical record of the rupture indications observed prior to and following the shutdown. The operating history of tube 3361 is compared with an adjacent tube, with no abnormal conditions observed at any time. Reactor temperature and power history since initial primary loop heatup are presented to show the total operating history experienced by N-Reactor.

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.

The hot-die-sized concept has been under investigation as an alternate manufacturing process for several years. Irradiation testing of the elements began in mid 1963. Some dimensional behavior problems were encountered with the fuel and a test is now being irradiated which may provide sufficient data to remedy the situation. Pre-irradiation examination of fuel elements has shown a significant dimensional change of the cores during the fabrication process, and this dimensional change is thought to contribute to the irradiation behavior of the elements. The irradiation effects of fabrication parameters that influence the dimensional. The objective of the test detailed in this is to evaluate the effects of end bonding pressure and machined core size on the irradiation characteristics of hot-die-sized diffusion bonded fuel elements.

A number of tests have been performed and reports issued concerning the adequacy of the last-ditch cooling systems at the Hanford small production reactors. At the present time, re-evaluations are being made, both theoretical and experimental, by process engineers and others at th particular reactor sites. In the interest of uniformity and consistency, this report presents a means of determining the last-ditch cooling adequacy for all the small production reactors. This method includes both the night tanks and the export system.

Two changes have been incorporated in the tables of this draft: (1) The lithium-aluminum tonnages required for blanket and E-N loadings have been changed, (2) The production numbers for the small production reactors have been reduced per AEC direction, and a separate table is included for these numbers. Included data are power levels, operating efficiency, exposure, tonnage, conversion ratios, and annual production for the C, K, and small reactors.

The horizontal control rods presently installed in the older Hanford Reactors have inadequate heat transfer characteristics for present and predicted future operation of the reactors. Continued graphite stack distortion, coupled with higher graphite temperatures, has resulted in ECR failure during reactor operation, such as swelling of the outer aluminum sheaths to the extent that rod movement in the graphite channel is severely restricted. Continued graphite stack distortion will tend to further aggravate the problem of rod operation. A new HCR design concept,was developed by P. H. Hutton of Reactor Design, IPD, to alleviate some of the pressing operational problems. Prior to the acceptance of such a design, the important rod operating parameters should be known to some degree of accuracy. This study was conducted to detexmine, by calculational methods, the temperature distributions that could be expected to occur in such an HCR when used at the present operating power levels and at 120% of the present power levels.

This report details activities of the Metallurgy Development Operation for the months of April, May, June 1964.

Quarterly report regarding the activities of the Yankee Core Evaluation Program, the purpose of which is to evaluate the performance of the core used in the Yankee Plant and to compare actual performance with predictions, as well as the revision of designs to address any differences between actual and predicted performance.

Abstract: This memorandum discusses a new design definition technique, requiring no standard, which pictorially shows part and gage limits on the same drawing. When functional gages, test fixtures, or optical chart gages are the selected inspection method, design, manufacturing process, and gage engineers can use these drawings directly to replace the usual standards-based drawings. In essence, this advanced technology visually describes true position dimensioning and tolerancing. It enables product designers to directly control the acceptance process.

Introduction: On November 14, 1961, an environmental test program commenced on the first SNAP 10A system structure, S10A-PSM-1.