This document details activities of the irradiation processing department during the month of July, 1958. 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.

Hanford fuel elements are made by brazing uranium cores in aluminum cans. The interface, or junction, between the AlSi braze and uranium core known as the compound layer, is composed of an intermetallic complex. The properties of this compound layer, brittleness, ductility, toughness, are considered to have a pronounced influence on the performance of fuel elements in the reactors. This report presents the results of initial efforts to identify the compound layers and establish some of the process conditions and variables which influence the properties of the compound layer, in particular, the lead preheat time.

The report is divided into: Fissionable materials (2000 program): studies related to production reactors, studies related to separations plants; reactor development (4000 program): Studies related to plutonium recycle program.

In the development of fuel elements for the NPR, one potentially serious fuel element problem -- high temperature uranium swelling -- has not been experimentally investigated. A series of experiments has been proposed in which uranium fuel rod with different amounts of Zircaloy-2 cladding will be irradiated to high exposure at temperatures equivalent to those expected in an NPR. These experiments should show the importance of high temperature uranium swelling as a limiting factor in NPR fuel element behavior. To obtain sample rod temperature of 250 to 300{degree}C on the surface and 500 to 650{degree}C at the center, the rods will be irradiated in aluminum capsules loaded in standard reactor process tubes. The high temperatures will be obtained by restricting the heat flow from the uranium sample to the coolant. The purpose of this test is to determine the validity of the heat transfer calculations used in predicting the temperature drops within the capsule by irradiating one capsule at known flux conditions and measuring the temperature attained by the uranium rod sample. The data obtained from this test will be used in determining the irradiation conditions required for the full scale uranium swelling tests.

This is the monthly report for the Hanford Laboratories Operation, July, 1958. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, employee relations, operations research and synthesis operation, programming, radiation protection, and laboratory auxiliaries operation area discussed.

The purpose of this project is to provide instrumentation at B, C, D, F, H,, KE, and KW Reactors that will (1) indicate front-to-rear differential pressure to assure process water flow in each group of tubes fed by a crossheader, and (2) actuate an alarm in the event of unsafe low flow conditions. The scope of this work consists of the installation of tubing from the rear and front crossheaders to differential pressure instruments on the front face; the provision of a signal panel and alarm horn on the front elevator; and the provision of a lead to the annunciator in the control room.

The Objectives of this test are to: (1) Investigate variables which may be responsible for the severe pitting of X-8001 Alloy jackets, and (2) Extend the testing of reduced pH coolant.

The July, 1958 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

This report details a meeting held for the purpose of acquainting all interested parties with the present problems being encountered in the canning process. A short discussion of previous canning methods and their problems was presented as background information. The search for a bonding agent to bond the aluminum can to the uranium core covered such materials as AlSi, lead, tin, aluminum, and copper. Mach of this early work was guided by the University of Chicago. The Grasselli (du Pont) Laboratory developed the triple-dip process where the uranium was first heated in a copper-tin bath, rinsed in a tin bath, then canned in an AlSi bath with AlSi as the bonding agent. It appeared at this time that the best AlSi composition to use was a little toward the aluminum side of the eutectic composition of 12.5 percent silicon. A current problem and the one with which this meeting was primarily concerned is that of bond failure where the bond between the can and core is broken. Variables found in this most recent study to have an effect on the strength or quality of the bonding layer are discussed.

This investigation is the second study of short period flow decay in the Hanford reactors. The first study was made on the D-Reactor. As in the analysis for D-Reactor quantitative analyses were made for ruptures that might result from sabotage. A visual inspection of the entire process water piping system was also made to check for any glaring deficiencies with respect to earthquake resistance. A brief discussion of the earthquake resistance of the system is also included in this report.

A direct means of monitoring and indicating the neutron flux in a reactor core is required to maintain neutron-flux equilibrium and to eliminate hot spots, flux peaking and flux oscillations. This report presents operating specification for an in-core flux monitor. The specifications are based upon the recognized requirements and operating characteristics for Hanford reactors and are presents as the criteria for development, application, and evaluation of prototype in-core flux monitor systems.

Reactor operating conditions are formally established to conserve pile life. A major part of the 2000 Research and Development effort is oriented to develop fundamental information required to properly establish these limiting operating conditions. This report reviews the current condition of reactor components believed important to continued productive life of the reactors and discusses the current status of technologies important in estimating the future productive life of the reactor complex. These evaluations are conducted continuously and this report is one of a series of status reports.

None

Experiments were performed with 2.09% U/sup 235/-enriched uranium metal plates in a light-water-moderated and -reflected assembly. Each plate was 30 in. long, 31/8 in. wide, and 1/4 in. thick and contained 7.09 kg of uranium. In the first assemblies the plates were arranged in rows with edges adjacent, and the spacing between rows was varied from 5/8 to 1 1/8in. The optimum spacing was about 7/8 in. With the spacing between rows maintained at 7/8 in., the spacing between the edges of the plates was varied from 0 to 5/8 in. The optimum spacing between edges was 5/16 in. The minimum critical mass in these assemblies was 6.74 kg of U/sup 235/. (auth)

The effect of hot-worked thickness on the corrosion behavior of atmosphere-melted Zircaloy-2 is discussed. Hydrogen pickup upon etching Zircaloy- 2 Zircaloy-3, and properties of vacuum-cast Zircaloy-2 and the fabrication of special Zirccaloy shapes are presented. A hardness converison chart for annealed zirconium and zirconium alloys is included. (J.E.D.) The useful properties of the tungsten wires used in the incandescent lamp and ra dio tubes industry are ensured by adding about 1% of various ingredie (e.g., K, Na, Al, Si compounds) to the tungstic acid and producing metal powder and compressed rods. During the sintering of these rods, the additions are evapo rated to such an extact that finally the concentration of foreign atoms remains only 10/sup 5/. According to the investigations, the warm hardness values of these tungsten pieces is determined by the traces of impurities. In order to investigate the influencce of the various kinds of atoms, an apparatus was built for the determination of half-micro warm hardnesses of various kinds of tungsten metals in the temperature interval 20 to 800 deg C The tungsten wires made with additions containing Si had an average hardness H/sub V/ = 250 at 800 deg C, while the wires without such additions …

The effects of beams of protons and alpha particles from the 60-inch cyclotron and beams of accelerated nuclei of heliuan-4, carbon-12, oxygen-16, and neon-20 from the linear accelerator on thin films of dried enzymes were measured. The experimental apparatus is described and results are presented graphically. (C.H.)

Extensive studies have been conducted of kerosenetype diluents for the tri-n-butyl phosphate (TBP)kerosene solvent used in uranium extraction. Wide variations have been shown in the paraffinic, isoparaffinics aromatics and olefinic hydrocarbon contents of these diluents. Such diluents have been shown to be unstable in thc presence of high nitric acid concentrations, and in the presence of small amounts of nitrous acid. This instability is primarily a function of the non-paraffinic hydrocarbon content of the diluent and of the nitrous acid concentration of the contacting medium. Degradation of a diluent has been shown to discolor a diluent, increase its density and viscositys and influence uranium distribution during re-extraction operations. Solvent treatment with Na/sub 2/CO/sub 3/, followed by H/sub 2/O and nitric acid washes, has been shown to be only partially effective in removing diluent degradation products. The diluent degradation products have been characterizeds chemically and spectrallys as carboxylic acids, organic nitrates, organic nitro compounds, and nitroso oompounds. (auth)

In a study of the nuclear and physical properties of the concrete shield of the ORNL Graphite Reactor measurements were made of the fast-neutron fluxes in a hole through the shield. These fluxes were not incinded in the formal report of the study (ORNL-2195) since several uncertainties associated with the measurements made the absolute values unreliable. The attenuation curves at one energy may be reasonably accurate, however. (auth)

Speciments of iodide zirconium, Zircaloy-2, Zr-15Nb, iodide titanium, TMCA-45 titanium, A-110AT titanium, and 430 stainless steel were corroded in a highpressure recombiner loop. Analyses were performed to determine the amount of hydrogen pickup. The titanium materials and iodide zirconium showed very high hydrogen pickups, while the zirconium alloys and the 430 stainless steel absorbed smaller amounts of hydrogen Metallographic examination of the specimens showed that recrystallization occurred in all but the Ar-15Nb specimens. There seems to be little difference in the extent of recrystallization and grain growth whether the in the recombiner section at 430 to 500 deg C. Recrystalliplained or correlated in any way with the amount of f hydrogen sion that occured. Since hydrogen is known to seriously embrittle zirconium and titanium, it is recommended that crystal-bar zirconium and titanium alloys not be used as materials of construction in environ ments sinmilar to that of the High Pressure Recombiner Loop. (auth)

Two problems are encountered in U/sup 233/b purification and metal production. One is high alpha activity, which requires that all operations be enclosed. The other is beta-gamma activity originating from U/sup 232/ present in U/sup 233/ as a contaminant. This, the immediate daughter of U/sup 232, produces a chain of strong beta-gamma emitters, which make shielding and thorium separation procedures necessary. Chemically impure solutions of metal fabrication residues are separated from trace amounts of thorium isotopes by cation resin exchange in shielded equipment. The short-lived thorium daugters are allowed to decay, after which beta-gamma activity is low enough to permit chemical purification in unshielded equipment. The U/sup 233/is chemically separated from nonradioactive contaminants by precipitation as UO/sub 4/- xH/sub 2/O, which is calcined to U/sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is reduced to U0/sub 2/ and converted to UF/sub 4/. Pure metal is prepared by bomb reduction of the UF/sub 4/ with calcium as the reducing agent. Residues from the chemical purification steps, reduction, metal fabrication, and ganeral cleanup are processed into nitrate solutions, solvent extracted, and concentrated for recycle. Ion exchange, precipitation, reduction to metal, and casting all have effect upon the thorium isotopes and their daughters, resulting …

A study was made to determine the magnitude of the velocities existing in the various slurry pumps currently in use and to determine possible methods of reducing the velocities in order to obtain longer life. The maximum velocities of the lOOA, 200A, and 300A pumps are between 39 and 53.6 fps with the exception of the lOOA absolute discharge velocity which is 80.6 fps. For large slurry pumps, it is recommended that lower speeds be used in order to reduce the impeller relative velocities. (auth)

Experiments are described which studied the effect of irradiation upon small U/sup 235/ cylinders simulating in size those designed for thc initial CP-4 (EBR-I) fuel loading. Qualitative similarities between thermal cycling and irradiation effects were found, but none of the three types of inaterial tested were entirely dimensionally stable. (auth)

The feasibility of utilizing a jet pump and a shroud as a recirculation device within the blanket of the proposed reactcr pressure vessel for the HHE-3 was examined A Schute-Koerting designed jet pump was tested in conjunction with a 1/3 scale plastic model consisting of a core vessel, two constant area shrouds, and a blanket vessel. It was concluded that a jet pump with a ;multinozzle in conjunction with a shroud designed with a sizeable annular flow area provides a very satisfactory recirculation device within the blanket of a two-region reactor pressure vessel. A larger circulating pump head is required, but the fluid flow in the external circuit may be reduced consideraly depending on thee shroud used. (W.D.M.)

Experiments were performed to determine the critical parameters of aqueous solutions of 93.2% U/sup 235/-enriched uranyl fluoride contained in cylindrical annuli formed by various combinations of aluminum cylinders varying in diameter from 2 to 30 in. In all of the experiments the inside cylinder was lined with a 20-mil-thick cadmium sheet and filled with water to a height of 48 in., and in some experiments a water refiector was used on the sides and bottom of the outside cylinder. The data indicate that for the solution having an H:U/ sup 235/ atomic ratio of 50.4 the critical infinitely high reflected annulus would have a minimum thickness between 2.5 and 3 in., while the unreflected annulus would have a thickness between 3.75 and 4.5 in. The corresponding thicknesses for the solution having an H:U/sup 235/ atomic ratio of 309 would be between 3.5 and 4 in. for the refiected annulus and between 4.5 and 5.5 in. for the unreflected annulus. (auth)