The average energy release per fission event in a reactor is dependent on the composition and arrangement of the lattice materials. In a study of heat generation in the NPR, Nilson developed expressions for calculating the average energy released in each material per fission event. These relationships have been used in the present calculations to obtain the energy release per fission in existing Hanford reactors.

Irradiation Processing Department opinions regarding HCR sleeves. It is believed that Zr-2 would be a satisfactory sleeve material providing it is limited to maximum service temperatures the order of 400 degrees C. Both Inconel and a 300 series stainless steel are believed possibilities for uncooled sleeve materials, with Inconel our preference. Beryllium, in virtue of its outstanding nuclear properties and probably satisfactory chemical and mechanical properties, warrants consideration on a pilot basis if fabrication of a suitable sleeve form is practicable. Consideration of any sleeve material should include techniques for minimizing the friction between the sleeve and the control rod -- particularly when a fretting -- prone material, such as Zr-2, is involved. A ball, or roller, bearing rod support, with a ceramic bearing contacting the sleeve, might be considered as a device for reducing sleeve wear and improving rod manipulation.

This monthly document details activities of the Hanford Operations Office during the month of April 1960. (FI)

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The purpose of this document is to establish the design scope for the proposed modifications to the reactor ``last ditch`` cooling systems in the 100-B, D, F, H, DR, and C Areas. The objective in making these modifications is to provide adequate ``last-ditch`` reactor coolant flows for safety of operation at power levels currently programmed for the period CY 1964 when additional ``last-ditch`` cooling facilities are planned in connection with major plant modifications. Additional interim modifications may be required for the last ditch system at the 100-C and DR Areas and for the export water system prior to major plant modifications during CY 1964--1965.

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The objective of this supplement is to authorize charging of additional dingot uranium and to extend the charging to all old reactors, (B, D, DR, F and H). The original test authorized charging 360 tons of dingot uranium at a rate of approximately 60 tons per month. Charging began in May 1960 and at present the majority of this material has been charged. The present delivery schedule calls for accepting 90 tons of dingot uranium per month from MCW, starting in November 1960. To meet this schedule it is necessary to either authorize additional charging under a suitable test program, such as this PITA, or to accept dingot uranium as normal material.

This document details activities of the irradiation processing department during the month of August, 1960. 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.

A nomogram is attached as Figure 1 which may be used to determine TAI limits. The nomogram was developed to meet the need for a rapid and accurate method for calculating these limits. The technical basis for determining individual tube outlet temperature limits is given in HW-65729.

A resume of Health Physics activities for April, May, and June, 1960 is presented. Discussions and tabulations which summarize results of field surveys, bioassay, personnel monitoring, and environmental surveys are included.

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Heat-transfer data are presented for the Experimental Gas Cooled Reactor Title I seven-rod fuel-assembly design. The effect on heat transfer of (1) the radial location of the outer six rods of the seven-fuel-rod cluster and of (2) the addition of helical-finned spacers at the midpoint of each of the seven fuel rods is discussed. The heattransfer data were obtained to verify preliminary general assumptions pertaining to the heat-transfer characteristics of the seven- rod fuel-assembly design and to obtain local heat-transfer correlations. The heat-transfer tests were performed at near-atmospheric pressure using air as the coolant medium. Plots and equations of heattransfer correlations over a Reynolds Number range from 12,000 to 80,000 are included. The test set-up and test procedure are also described. (auth)

This specification covers the design, fabrication, testing and furnishing of two fire pumps complete with drivers and appurtenances.

This document details activities of the irradiation processing department during the month of July, 1960. 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.

This document details activities of the irradiation processing department during the month of September, 1960. 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.

A preliminary flowsheet for the purification of uranium dioxide fuels by a magnesium reduction-- mercury extraction-- steam oxidation process is proposed. Feasibility was indicated by laboratory-scale scouting experiments. Data evaluation indicated 100% reduction of uranium dioxide by magnesium although this figure was not demonstrated, chiefly because of poor choice of materials and design of equipment. Steam oxidation of uranlum tetramercuride produced an oxide with an O/U ratio of 2.43. This ratio was decreased to 2.09 by heating the oxide in a hydrogen atmosphere at 900 deg C for 1 hr. The final product had a surface area of 3.5 m/sup 2//g, and 18% of the panticles were < 1 mu diam. A pellet of the oxide sintered at 1750 deg C had a density of 9.76 g/cc, 89% of theoretical. Decontamination factors demonstrated for ruthenium, cesium, and samarium, when present originally in amounts equivalent to 30,000 Mwd/ton fuel burnup and 60 days' decay, were

A process was developed for separating promethium from raixed fisBion product rare earths by continuous multistage conntercurrent extraction with 100% tri-nbutylphosphste from nitric acid of 12 N or higher concentration. Distribution coefficients at 12 N acidity for aecdamium. promethium. and samarium are 0.43. 0.82, and 1.55, respectively. Single-stage separation factors of 1.9 between successive elements can be maintained throughout the system to give separations dependent only on the number of stages. Extracted values can be recovered from the organic solution by stripping with a smaller volume of dilute nitric acid. A flowsheet for purification of promethium includes one cycle for separation of promethium from neodymum and lighter elements and a secondycle for removal of samarium and heavier elements. Each cycle consists of a series of countercurrent partitioning stages. followed by stripping stages and an evaporator. With 20 stages in the first cycle and 34 stages in the second, a 90% yield of promethium with a purity of 83% can be obtained from a typical mixture of fission product rare earths, assuming essentially perfect mechanical efficiency. An increase to 34 stages in the first cycle would permit a 93% yield of 99% promethium. (auth)

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An economical process was successfully demonstrated in bench-scale continuous equipment for stripping U from amines with ammonium carbonate solution. A continuous countercurrent mixer-settler extraction system was set up for further testing of the process for recovery of Te, Np, and U by tertiary amine extraction from UF/sub 6/ transfer cylinder was solutions. The effect of Purex aqueous feed adjustment procedures on Pu extraction by 1 M di-secbutyl phenylphosphonate (DSBPP) was studied. Work was continued on plutonium(IV) nitrate extraction with TBP and phenylphosphonate esters. The response of Ru/sup 106/ extraction to variations in the treatment of TBP-Amsco 125-82 solvent was tested. Two solvents have shown ability to extract cesium. (For preceding period see CF-80-3-136.) (W.L.H.)

A solid-state bonding technique involving the use of gas pressure at elevated temperatures was utilized for the self-bonding of molybdenum and niobium. Bonding conditions and surface preparation as a function of the integrity of the bond achieved were evaluated for each material. Optimum self-bonding of niobium was achieved by bonding parameters of 2100 to 2300 deg F at 10,000 psi for 3 hr with surfaces which had been prepared by etching in a nitrichydrofluoric acid solution prior to bonding. The process as developed was used to prepare niobium- clad flat-plate- and rod-type fuel elements and flat-plate subassemblies. Niobium tubing was also fabricated by this technique. (Molyb denum self-bonding was most readily achieved by gaspressure bonding at temperatures of 2300 to 2600 deg F at 10,000 psi for periods of 3 hr. With these bonding conditions a number of different surface preparations were satisfactory. Directional ductility of the molybdenum was encountered after bonding and methods to eliminate this were evaluated. Cross rolling with respect to the original rolling direction was shown to improve the ductility of molybdenum-clad specimens. (auth)

The effect of two neutron poisons, baron and cadmium, on the rate of dissolution of high-density 95% ThO/sub 2/-5% UO/sub 2/ pellets in the Zirflex Process was determined. Dissolution of U-10% Mo alloy in boiling HNO/sub 3/ resulted in a precipitation of uranyl molybdates. Air caused greater uranium and thorium losses during decladding of ThO/sub 2/-UO/sub 2/ fuel than irradiation. Processing of U-Mo fuel by a Zircex type process is discussed. Two leaches of graphitized fuel with 90% HNO/sub 3/ recovered more than 99% of the uranium. Irradiation of synthetic ThO/sub 2/-UO/sub 2/ fuel solution to 5 and 10 watt-hr/l in a Co/sup 60/ source resulted in about a 50% decrease in decontamination factor using the acid-Thorex flowsheet. Corrosion of titanium, tantalum, and Ni-o-nel in Thorex solution and titanium corrosion in various molybdenum core alloy solutions were investigated. The solubilities of ferric mono- and dibutyl phosphates in HNO/sub 3/ and 30% TBP-Amsco-HNO/sub 3/ solutions were determined. Fission product concentrations expected in Purex waste from processing Yankee Atomic Reactor fuel were calculated. Chemical applications of nuclear explosions to H/sup 3/ exchange, reduction of CaSO/sub 4/, and Gnome sampling are discussed. (For preceding period see CF-60-6-108.) (M.C.G.)

SPERT I: The characteristics of the boiling process and its relatin to moderator expulsion in Spert I were investigated in a series of capsule type experiments. A fuel-bearing oapsule, instrumented to provide pressure, volume, and temperature data during transient power excursions, was placed in a high flux region of the Spert I P core. Five step-induced transients initiated from boiling indicate that the kinetic behavior of the stainless steel clad P-18/19 core is dependent on initial temperature in a manner similar to that of previously tested aluminum clad spert cores. Reactivity oscillator techniques were used in the P-18/19 core to determine the phase and magnitude of the reactivity-to-power transfer function from 0.01 to 18.4 cps at low power and at temperatures below boiling. Criticality data on relatively simple lattices, both rod-free and containing a single poison rod, were obtained from a series of clean critical experiments performed on a number of light water-moderated and - reflected slab configurations of Spert III fuel elements. Changes in water height during the critical water height experiment were measured to plus or minus 0.0013 inches by means of a simple remoteindicating system designed and built for this purpose. SPERT III: The operational loading was …

>A handbook containing all available current physicochemical data on UF/ sub 5/ is presented. Every effort was made to obtain and consider all reports of original data for incorporation in the compilation. One hundred and forty nine references are given. (J.R.D.)

The external holdup of expensive materials and associated capital costs for the heat removal systems of fluid fuel breeders were determined. The aqueous homogeneous and molten salt breeders were found to contain substantially less uranium holdup external to the core than the liquid metal fueled breeder. The cost of heat removal and turbogenerator plant equipment for the three systems was compared. (auth)