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This document details activities of the Reactor Branch during the months of January through December 1963. (FI)

About 298,000 OIIIN aluminum components were received from Harvey for evaluation after major steps were take in the vendor`s plant to upgrade product quality. This material met HAPO specifications and was comparable to Alcoa control components. Finished fuel elements (non-bumper and bumper) canned in the test and control components are tabulated by uranium core type and canning month. Non-destructive and destructive test data did not show any differences in clad fuel elements attributable to component quality. Irradiation status of non-bumper and bumper fuel elements clad in Harvey improved components as of Dec. 31, 1962, is tabulated, as is a summary of ruptures. It was recommended that Harvey Aluminum Co. be allowed to bit competitively on all future Al component orders except for the KVN and KVE component models.

These tests were performed in accordance with a letter fran R. C. Breisch to J. R. Spink, May 7, 1962. An unrestrained gunbarrel condition was simulated for each of the standard process tube inlet assemblies for B,C and K type reactors. Primary objects of the testing were to obtain data on movement, pigtail damage, and effect on flow to the process tube in case of a Van Stone flange failure without key restraint of the gunbarrel. All test conditions were relative to actual conditions, but these conditions were not necessarily typical or maximum. The gunbarrels were not loaded with do-nuts or thermal shield pieces. All of the assemblies moved instantly and rapidly when released under pressure. The B and K braided type pigtails crimped just upstream of the ferrule on the nozzle connection. The helical tube C-type pigtail stretched about six inches, but did not travel far enough to cause flow restriction. Tube flow for the B case was 54 per cent of that prior to Van Stone flange failure. For the K test, the flow was reduced to 69 per cent of that prior to Van Stone flange failure.

This document has been prepared for internal use by the General Electric Company to serve as a program for evaluating reactor instrumentation and safety circuit equipment needs. It is intended that this document be used as a guide for defining, planning and scheduling engineering effort; budgeting of capital money; and project planning for new instrumentation systems. Effort will be made to periodically evaluate the status of the programs presented and provide updating information accordingly.After a plant has been built and operated for a number of years, it becomes apparent to operating and engineering personnel that certain modifications in controls and monitoring systems would provide both tangible and intangible benefits. Systems which were once thought to be the primary points of control shift in importance as others become recognized. As time passes this shifting spreads the main control focus from the central control desk to various other areas in the control room. Production rate increases cause instrument ranges and scales to be changed so that information on the process can still be obtained from existing equipment. Response times, sensitivity, limits, and time constants which were figured for one level must be used or revised for new levels. Further, it is discovered …

The objectives of this production test are to confirm the KVNS fuel element design, determine the effect of support height on annulus coolant temperature distribution (R Value), evaluate the effect of eccentricity on KVNS fuel element irradiation behavior.

This document summarizes the graphite treatment studies conducted under Design Test 1136 in connection with the retubing of the K Reactors with zirconium tubes.

A half-plant low alum water treatment test began at F Reactor on January 16, 1963 at startup from the scheduled January 3 tube replacement outage. The test, which was prompted by results obtained from a statistical analysis of fuel ledge corrosion attack, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from normal production fuel irradiated in two different alum treated process waters. This report discusses the results obtained from twenty fuel charges, ten from each side of F Reactor, which were discharged prior to the reduction in alum feed to establish the pre-test corrosion environment.

This is the monthly report for the Hanford Laboratories Operation January 1963. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

This report details test plans for production test IP-560-A. The objective of which is to determine the corrosion effects resulting from reducing the amount of sodium dichromate corrosion inhibitor added to process water from 1.8 to 1.0 ppm at a process water pH of 6.6.

This report, for January, 1963 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; finished products operation; financial operations; facilities engineering; research employee relations; weapons manufacturing operation; and power and several maintenance operation.

A half-plant high flocculation pH test began at B Reactor on September 24, 1963. The purpose of the test was to determine whether operation of the water plant flocculation basins and filters at a pH higher than 7.0 with subsequent coolant pH adjustment to 6.6 is beneficial as far as radioactivity in the effluent is concerned. It is planned to increase the flocculation basin pH in steps to a maximum of 7.6. Although primarily an effluent activity test, fuel discharged after a metal cycle at each water plant condition will be monitored by visual examination. Thus, it will be possible to determine whether the incidence of localized corrosion is influenced by the various modes of operation. One 20 column fuel discharge, ten columns from each side of the reactor, was obtained prior to the start of the test. This report discusses the results obtained from the first discharge of fuel under test conditions.

The discharge of radioactivity from Hanford production reactors to the Columbia River has been studied for a number of years. The primary purposes of these studied have been to learn of the factors influencing production reactor effluent radioactivity concentrations and to develop methods for lowering the discharge concentrations of these radionuclides. In 1962 the Coolant Systems Development Operation (CSDO) began work on this program to determine the in-reactor film compositions and the inventory of radionuclides held in these films with the long-term objective of learning more about the complex in-reactor retention reactions that influence radionuclide effluent concentrations.

The real cost of metallurgical examination of irradiated fuel elements is several times that of fuels before irradiation. Since post-irradiation study is necessary to determine the performance of prototype fuel designs in a reactor environment, complete physical characterization before irradiation provides an economical and efficient basis for comparing experimental data. Without such comparative study, unexpected damage to fuels in the reactor would be difficult, if not impossible, to interpret. This report presents such a characterization of one fuel element from a group of fuels prototypic of the initial charge in Hanford`s N-Reactor. This fuel element, 1055-8, is significant because it belongs to the Production Work Request CX-2 family of elements.

The control rod calibrations in the E-N core loading and in the natural uranium core with an E-N blanket yielded results which indicate that the controls are approximately 10 per cent less effective in an E-N core due to shadowing of the rod strength by the E-N lattice; it had been expected from theoretical, calculations that the E-N shadowing effect would be approximately 10 per cent. The tests thus indicate that there has been approximately 25 per effective rod calibration value since the 1957 calibration due to changes in the flux pattern associated with alterations in fuel type and geometry. The significant difference between the measurement made in 1957 when the loading was basically natural uranium, and the measurement made in November 196l, when the pile contained E-N fuel, is thus concluded to be a combination of the blacker lattice and flux distribution effects.

During the 18-year history of Hanford operations, a high degree of safety has been provided at reasonable cost by the storage of high-activity liquid wastes in underground tanks, by the percolation of low-activity liquid wastes (principally water) through the soil to the ground water, and by the adsorption or filtration of radioactive materials from gaseous effluents prior to their discharge to the atmosphere. Studies of the Hanford area indicate that current practices could be continued almost indefinitely without jeopardizing environmental safety, provided the stored liquid wastes are periodically transferred to new tanks before the existing ones fail. In 1960, a Fission Product Recovery Program was prepared outlining a plan to meet the Atomic Energy Commission`s needs for isolated fission products. Emphasis was placed on developing processes for isolating selected fission products and on providing interim production capability for recovering the fission products of immediate interest. In 1961, development efforts were channeled almost exclusively toward meeting the immediate needs of the Commission for separated strontium-90 and cesium-137. During this time, it became increasingly evident that substantial economies could be realized, both in immediate development and subsequent operating efforts, if the separate fission product recovery and waste management programs were merged. The …

This report details activities of the N-Reactor Department during the month of January 1963.

This document details activities of the irradiation processing department during the month of January, 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.

A new, dry process for the production of UO/sub 2/ powder from UF/sub 6/ by fluidization and particle-coating techniques was developed. The process involves direct conversion of UF/sub 6/ to a solid, UO/sub 2/F/sub 2/, by hydrolysis with steam and the reduction of the UO/sub 2/F/sub 2/ to the dioxide by reaction with was carried out in 3-in.-dia. Monel reactors. In continuing studies, dense, spheroidal UO/sub 2/ particles in the range --20 +200 mesh are being prepared. A simultaneous (steam and H/sub 2/) reaction procedure at temperatures up to 7OO deg C is being used. Particle densities ranged to 9.75g/ cc or about 89% of theoretical. Sintering in H/sub 2/ at about 1700 deg C for 2.5 hr gave final densities of 10.4 to 16.6 g/cc or up to about 96.5% of theoretical density. (auth)

Discussion of theories on particle motion in electric and magnetic fields precedes examination of theories on the source of the inner and outer Van Allen belts. Theories for the inner belt propose solar cosmic ray injection and galactic cosmic ray injection as the source of the protons. Neutron albedo is particularly examined but seems to be unaccountable for the proton belt. Neutron albedo and injection from the sun are also considered as possible sources of the electrons in the outer belt. Fluctuations, intensities, and distributions of the belts are also discussed, and the connection between the belts and aurora displays is examined. Fluxes, spectra, and particle intensities are displayed, and the effects of magnetic storms on the belts are analyzed. 150 references. (D.C.W.)

A system of six codes is described that permits the solution of a broad class of problems encountered in the study of neutron thermalization, average cross sections, and spectral models. These programs are all compatible with use on MONITOR for the IBM 7090. The Fortran listings are given. (auth)