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Production of Pu nitrate from separations plants during May was below forecast. A Np recovery campaign in Purex yielded 1.5 kg. Production and shipments of UO{sub 3} met schedules. Unfabricated Pu metal production was below forecast, but all shipments were on schedule. Decontamination efficiency was low in Purex solvent extraction around the time of the Np recovery. The damaged Redox B-2 dissolver is being restored; processing of enriched metal in A and C dissolvers was continued. A spectrograph for inclusions in Pu metal was installed. 4 kg Pu oxide was produced in a continuous direct calciner. Scope design on Purex Np recovery and purification facilities was completed. Other design and contracts are discussed.

This report, from the Chemical Processing Department at HAPO for May 1966, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee-relations, and waste management.

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From introductory paragraphs: "The central problems in neutron research are the understanding of nuclear structure and the study of the properties of nuclear particles, particularly the properties of the neutron. The most fruitful attack on these problems is the determination of the probability of interactions between neutrons and nuclei, i.e., the measurement of neutron cross sections. Ideally, this involves the study of all possible types of neutron interaction with all available nuclei at all neutron energies...The discussion in this paper will omit the interactions leading to neutron productions, and will be limited to the intersections of neutrons with stable nuclei."

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A two-foot-square fluidized bed calciner was designed and operated to convert aqueous, highly radioactive wastes into granular solids. The calciner exceeded its designed feed capacity of calcining, at a bed temperature of 400/sup 0/C, 100 liters per hour of aluminum nitrate solution simulating wastes from the reprocessing of spent aluminum-uranium alloy reactor fuel. Heat was supplied to the calciner by circulating NaK with an electromagnetic pump, at temperatures up to 1400/sup 0/F, through a tubular heat exchanger placed directly in the fluidized bed of solids. The results of ten runs are presented and discussed. Equipment development progressed to the point where a continuous, trouble-free, operating period of 43 days was attained. Several properties of the alumina product were routinely measured, and some of the effects of the calciner operating variables on these properties were determined. The production of both amorphous and alpha crystalline material was found to be possible; the crystalline nature of the product had a profound effect on product properties and off-gas loading.

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The Expansion Joint Test are detailed in this report are part of the work accomplished under Design, Development and Research Contract DDR-111 between General Electric Company, Hanford Atomic Products Operation and Washington State University. The equipment and instrumentation used for the K-Downcomer tests were arranged to permit installation and testing of the expansion joints.

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Interim results of analytical investigation of nuclear and thermal characteristics of SM-2 type fuel elements in SM-1, SM-lA, and PM-2Ae reactor cores are reported. Utilizing modified two-proup diffusion theory, predictions of power distribution and core and rod reactivity were performed. The calculations indicate that use of SM-2 fuel elemente in SM-1, SM-lA, and PM-2A is feasible from the nuclear standpoint. A steady state thermal analysis of each plant utilizing SM-2 elements was carried out. This analysis showed that the minimum departure from nuclear boiling ratios were considerably abovs the minimum value ftom the design criteria standpoint at both the operating and scram power levels. The investigation indicated that SM-2 elements can successfully be employed in SM-1, SM-lA, and PM-2A as replacement core elements. (auth)

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

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

Measurements were made of the air leakage rates through structural components that penetrate semicontainment reactor housing installations. Full- sized test specimens were sealed inside a 10-ft-diameter test sphere, and the leakage rate was measured as a function of pressure for pressure differentials of up to 25 in. of water. The data were fitted to empirical equations that describe the leak rate at any pressure differential within the range of the experimental tests. The forms of the equations are such that the total leak rate of a building can be obtained by summing the empirical constants and multiplying by a probable pressure differential. (auth)

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This is one of twenty-one volumes summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This portion describes work on Metallic Fuel Element Materials.

Mathematical relations are presented for estimating the release fractions of gaseous fission products from coated fuel particles and fuel elements containing them. The relations are based on simplified models of the release process, with particular emphasis on the following mechanisms: recoil, diffusion from fuel, diffusion through particle coating, and diffusion through fuel element matrix. The characteristics of fission-gas release by these mechanisms, acting singly and in combination, are considered, and the application of the theoretical relations to experiment planning and interpretation is discussed. Special attention is given to methods for analysis of data from continuous, in-pile (sweep capsule) release experiments and neutron-activation release experiments. (auth)

This memorandum discusses study areas and items that will require attention for the naval studies of the utilization of nuclear propulsion in a submarine-based missile system.

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Pilot plant studies were conducted to develop a continuous process for recovering uranium from nichrome fuels (HTRE). The process consisted of dissolution of fuel in mixed HCl--HNO/sub 3/ solution, removal of the chloride ion by stripping with HNO/sub 3/ in a packed column, and then recovery of the uranium by TBP solvent extraction. Recovery of uranium from nichrome fuels at satisfactory rates and efficiencies can be obtained by this process. Reactor fuels containing large amounts of silica may present a solids problem during chloride removal. Titanium is a suitable material of construction for the dissolver, stripping column, and their associated equipment (feed, off-gas, and product vessels). Stainless steel equipment is suitable for the solvent extraction system. (auth)

Report documenting a process for recovering uranium from nichrome fuels (HTRE). This is accomplished by dissipating the fuel into a mixture of HCl-HNO3, stripping away chlorine ions with HN03, and recovering the uranium via tributyl phosphate solvent extraction.

A superposition model is developed to predict the critical heat flux in forced convection flow. The model is applied to available experimental results in boiling water flows and good agreement is obtained between the model and test data over the multitude of geometries, flow rates, pressures, and fluid enthalpies tested to date. (auth)