A knowledge of the mechanical and physical properties of uranium is important in understanding its pile behavior. The effects of irradiation on the properties of uranium will be dealt with in another chapter. However, it is important to know and understand the pre-irradiation properties of uranium prior to investigating the effects of irradiation. This chapter discusses the mechanical and physical properties of uranium.

This report covers the examination of nine process tubes, five tubes from B Pile and four from H Pile. Four of the tubes from B File were removed because they were leaking and the fifth was removed because it contained a stuck ruptured slug. Tube 2681-H was removed for graphite mining, tube 2852-H was removed under PT 313-105-14M ``In-Pile Evaluation of 63-S Aluminum Process Tubes,`` and two tubes were removed from channel 0961-H under PT 105-506-E, ``Recirculation of Pile Cooling Water.`` Visual inspection was made of the inside and outside surfaces of the tubes before and after cleaning with cold 10 per cent nitric acid. Samples varying from one to four inches in length were taken from each section and transferred to the 108-B Metallurgical Laboratory for close examination to determine depth of pitting, 72-S cladding thickness, and wall thickness. These determinations were made at specific points in each section on what appeared to be the area of severest attack. It is indicated on six of the tubes where the sample was taken on the circumference of the tube.

A meeting of the Thorium Working Committee was held at FMPC on September 20, 1955. The current status of the thorium production program was discussed and a limited development and evaluation program was outlined which it is felt will (1) clear up some of the more important question marks in thorium metal quality and (2) develop a more satisfactory method for processing solid metal scrap. The basic parts of this immediate program are (1) SRL would evaluate ingot sections of as-cast double arc-melted virgin metal in order to classify typical ingot defects and determine their frequency of occurrence; (2) Du Pont would request for evaluation purposes ingot sections from typical recycle ingots which incorporated vacuum-induction melting for the consolidation of recycle metal in the form of consumable electrodes for arc-melting; (3) after completion of item (2), du Pont would then request approximately 200 slugs, made from suitably processed recycle metal, which would be used for canning and steam autoclave tests. This program should lead to definite specifications (and/or classifications) for ingots to be stockpiled for future AEC use.

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The isotope Na{sup 24} is one of the major short half-life constituents of reactor effluent water. It has a half-life of 15 hours and decays of emission of a 1.39 MeV maximum beta particle. Other radioisotopes present in reactor effluent water from which Na{sup 24} must be separated are Si{sup 31}, P{sup 32}, Mn{sup 56}, Cu{sup 64}, As{sup 76}, Sr{sup 91--92}, Ba{sup 139--140} and rare earths. This report describes a rapid and very simple method for the quantitative determination of Na{sup 24} in a radiochemically pure form.

The objective of this research is to determine the absolute entropies, heats of formation, and free energies of formation of selected salts in the temperature range of 800 to 1900 deg F. Assembling and testing of apparatus for measuring heat capacities from about 50 deg K to room temperature has been completed, including calibration of thermocouples by comparison with a platinum and with a copper resistance thermometer. This apparatus has been used to determine the absolute entropy of molybdenum hexafluoride vapor at 298.16 deg K. The result, 79.7 +- 0.6 calories/mole deg C, agrees well with that calculated from its spectrum, 80.05 cal/mole deg C. A solution calorimeter has also been assembled and tested. By comparing the heat of solution in aqueous sodium hydroxide of molybdenum hexafluoride with that of the oxide, the delta H of formation of the fluoride was found to be -392.2 cal/mole at 298.16 deg K. These data, together with other data extant in the literature, allow calculation of the desired thermodynamic functions over the temperature range of interest.

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A preliminary chemical flowsheet is presented of a fluoride volatility process for recovering and decontaminating U from heterogeneous reactor fuels after dissolution in a fused salt. In laboratory work, a gross BETA decontamination factor of >10/sup 4/ was obtained in the fluorination of a UF/sub 4/-- NaF-- Zr F/sub 4/ melt by passing the product UF/sub 6/ through NaF at 650 deg C. The solubility of UF/sub 6/ in molten NaF--ZrF/sub 4/ was shown in kinetic studies to cause a lag in the evolution of UF/sub 6/ from the fluorinator. Corrosion of nickel in the fluorination step appeared to be 2 to 4 mils/hr during the time that U was present. The average corrosion rate over the process as a whole yvas less than 0.4 mil/hr. (auth)

A list of studies approved at the meeting is given with programs outlined in the appendix. The following topics were discussed: (1) safety; (2) security violations; (3) arrangements made for interviewees with physics training to visit Building 777; (4) procedures written for handling various fissionable materials on a large scale in the fabrication laboratory; (5) backlog of orders at the Laboratory shop; (6) that the limited number of applications do not justify the use of depleted uranium in the Laboratory; (7) that Savannah River Plant can follow Hanford requirements as to the average and variation of U-235 content in sidestream materials; (8) that Laboratory is to supply the various dimensions and tolerances that probably will be required for tubes with spiral ribs; (9) shipping of various dies and aluminium shapes now being held by aluminium fabricators; (10) obtaining proper hydraulic performance of fuel assemblies; (11) preliminary scopes of work covering process requirements for producing natural uranium elements in plate and tubular form; and (12) arrangements to be made to review the Nuclear Metals Program.

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This is the weekly progress report for the 200 area for the week of November 10, 1955. Areas discussed pertaining to the solvent extraction plants include operation status, 232-F Building, 221-F - NFE dissolver, 200-F - solvent purification, 221-F - increased throughput, 241-F additional waste storage tanks, 221-F canyon pumps.

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