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)

One hundred and seventy five grams of americium in a hydrochloric acid solution varying from 1 to 7 N was converted to americium dioxide. Americium oxalate was precipitated from 0.1 N HCI with 100% excess oxalic acid and was converted to the dioxide by calcination at 800 ts C in air. The solubility losses in the oxalate precipitation filtrate averaged approximately 7 mg/liter of solution, with a total loss of 0.09%. (auth)

The maximum pressure rise in the core of the Homogeneous Reactor Test following a cold water accident during startup was calculated for upward flow of core fluid and for reversed flow. With the reactor initially critical at 260 tained C and a power level of 0.04 w, fuel solution at 100 tained C wse considered to enter the core at a flow rate of 150 gpm. With reversed flow this added reactivity at sn average rate of 1.7% o degradation t k/sub e// sec, while with upward flow the rate was 0.67% o degradation t k/sub e//sec. The core pressure incressed rapidly to a maxi-mum of 550 psi above normal operating pressure with reversed flow and 400 psi with upward flow when only the core pressurizer was available for fluid expansion. With the core and blanket pressurizers connected, the excess pressures were 375 and 210 psi, respectively. (auth)

The purpose of this test is to establish the reactor hydraulic flow vs pressure demand curve in the low-regions of around 4--10,000 gpm.

With the exception of a brief period of slightly elevated chloride level in the secondary blowdown, water-chemistry conditions during the period were satisfactory. During the period, the reactor was shut down for end-of-core-life testing and rearrangement. A set of specifications covering all electronic and electromechanical mechanisms required to control the SM-1 reactor through the rod- drive motors and clutches was prepared and issued. Installation of instrumentation for plant response and system performance was virtually completed. Work on the interpretation of long-lived radiochemical data obtained at the SM-1 during core lifetime was continued. Analysis of all fissionproduct data collected during Core I life has started. Thirty-eight stationary and seven control subassemblies from SM-1 Core II were checked for alpha contamination by a gas-flow proportional-counting technique. The work on the final design of a waste-disposal system for SM-lA was stopped and an investigation of an interim system containing a bypass sampling system was undertaken. Work continued on tests 202, 203, and 204 in the activitybuildup phase of Test Series 200. Core- physics measurements were taken at end of Core I life to complete the series of measurements made throughout the lifetime of the core. (W.L.H.)

Development of the SNAP I power conversion system is described. The system is designed to convert the thermal energy produced by the decay of radioisotopes into 500 watts of electrical energy by means of a mercury Funkine cycle. A list of specific accomplishments of the program is included. (J.R.D.)

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To determine the thermal stability of SM-2-welded plate type fuel elements, test specimens were subjected to temperature differences across plate width. Thermal deflections caused by the relatively cool side plates restraining the axial expansion of the fuel region were measured along the axial centerline of the test specimens. Region-averaged temperature differences varied from 0 to ll6/sup o/F, or about l35% of expected reactor operating differentials. Test specimens, machined from standard full-sized fuel elements, consisted of a single fuel plate and its proportionate share of element side plates, and displayed an l-shaped cross section. Thermal deflections of 0.005 in. maximum were measured at the expected reactor operating conditions of 87/sup o/F region- averaged temperature differences. With initial (cold) deflections assumed within the SM-2 tolerance of (?) 0.008 in., test results indicated that the total operating deflections will be (?) 0.013 in. maximum. (auth)

Progress on design studies under Task 3, the mechanical-design portion of the SM-2 core and vessel design, is summarized for the period Oct. 1955 to Mar. 1960. Task 3 covers the mechanical design of the reactor vessel, vessel closure, nozzle penetrations, steel reflector, core support structure, flow divider, control rods, absorbers, and fuel elements. Layouts showing the basic designs, major dimensions, and materials of construction are presented. Stresses for the reactor vessel selected were within ASME Code limits. The report does not contain final results of Task 3 work. (auth)

This report addresses the critical mass studies, part X.

Development of the control elements for the SNAP 1 power conversion system is described. A description of test and prototype hardware and performance data are included. The control package in its final design is a combination of regulator and speed-sensitive feedback which provides satisfactory steady-state operation and serves as a mechanism correction for system disturbances. (J.R.D.)

The solubllity of zirconium dibutyl phosphate in aqueous uranyl nitrate and nitric acid solutions was found to vary from 0.35 to 159 rng of zirconium per liter, increasing with increasing uranium concentration. The solubility in 1.126 M TBP in Amsco 125-82 equilibrated with uranyl nitrate-nitric acid aqueous solutions was found to vary from 0.33 to 3.55 g of zirconium per liter. Particularly in the exbactant solutions, the solubillty of zirconium as zirconium dibutyl phosphate is well above the zirconium content of extensively burned natural uranium fuels under Purex process conditions. The zirconium dibutyl phosphate studied was prepared by direct synthesis in aqueous solution and found to have a variable composition. Precipitation from 2 M HNO/sub 3/ at 40 deg C yielded a product which had the approximate composition Zr (OH)(NO/sub 3/)(DBP)/ sub 2/. (auth)

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Corrosion data for types 304L, 309SCb, and 347 stainless steel were obtained in a number of process vessels in the ORNL Thorex pilot plant during the development and the production-development periods of operation occurring between December 1954 and September 1956. Stressed corrosion-test specimens were exposed in the batch dissolver tank, the feed adjustment tank, the BT vapor separator, the A-column feed tuink, and the BTC catch tank Generally, types 304L and 3O9SCb stainless steel exhibited comparable corrosion resistance in all environments examined. Most of the studies were conducted with these two alloys. Severe corrosion damage was encountered in the vapor phase of both the batch dissolver twin, operated at a maximum temperature of 115 deg C, and the feed adjustment tank, operated at a maximum temperature of 155 deg C. Corrosion rates for types 304L and 309SCb stainless steel varied from approximately 30 to 55 mpy in the batch dissolver tank during the development and the production-development periods. Vapor-phase corrosion rates in the feed adjustment tank during the lant haif of the development period ranged from 85 to 100 mpy. Severe corrosion attack was experienced also in the BT vapor separator, which operated at a maximum temperature of 115 deg …

Twenty-nine experimental determinations of burn-out heat flux were made with water flowing by natural circulaion through electrically heated vertical tubes with and without internal twisted tapes and through rectangular cross sections of three aspect ratios. Heated lengths varied from 10 to 33 in., system pressure at the testsection flow exit from 14.7 to 26.3 psia, inlet subcooling from 36 to 170 deg F, and burn-out heat flux from 13,000 to 218,500 Btu/hr/sq ft. Tests were made with both unrestricted and restricted return flow paths. Three correlations were developed for predicting natural-circulation burn-out heat fluxes for such conditions. Two are useful for rapid estimation but the third involves a more fundamental assessment of the coolant mass velocity at burn-out by a graphical matching of the heat flux that a given flow rate can sustain to the heat flux that will produce that flow rate. For all the data, this approach gave average and maximum deviations of 15 and 38%, respectively. It was found that use of a slip ratio of unity is adequate for burnout prediction, and the reasons for this are discussed in detail. The small burn-out penalty incurred by a substantial restriction of return flow path, experimentally observed, is …

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This production test is designed to demonstrate that the K Reactors can be operated. with a higher graphite temperature limit than stipulated in present standards without a significant increase in the rate of either burnout or contraction of the graphite moderator stack. It is intended that the increase in allowable maximum graphite temperature will be utilized to permit operation with slightly higher reactor power levels with substantially the same enrichment level of the total uranium fuel charge as is currently in practice at the K Reactors.

A study was made to establish the validity of using homogeneous-lattice machine program calculations of the relative conversion ratio of the overbored lattice as compared to the present case. It is possible to draw both qualitative and quantitative conclusions from hand calculations utilizing simplified geometries; it is the purpose of this document to summarize such a study. Numbers included herein are not intended to represent absolute quantities, since the required calculations lie outside the scope of this study. Nevertheless, it is felt that a comparison of the overbored to the existing lattice based on the values presented here is a good ``first order`` approximation.

The development of rapid, selective, and sensitive methods for the determination of osmium in liquid samples is reported. The specific application of primary interest was the estimation of microgram or milligram quantities of osmium in homogeneous reactcr fuel solutions which contain uranyl sulfate as the major component and corrosion products of stainless steel as minor components. (W.L.H.)

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.

The objective of this test is to determine, through in-reactor testing, the resistance to corrosion of nickel plated fuel elements, plated by a chemical deposition technique. This program may eventually be composed of the following portions, the first of which is authorized by this test: (1) Irradiate ten columns of 0.5 mil chemically nickel-plated C-64 OIIN fuel elements having various heat treatments, each alternated with x-8001 control pieces. Two columns will be exposed to 400 MWD/T and eight to 800 NWD/T. (2) Should results from the above test be encouraging, separate authorization for the following tests may be requested: (a) Irradiate four columns of nickel-plated fuel and four of X-8001 clad fuel, both groups having purposely cocked pieces, until two ruptures are sustained in each group or until a factor of improvements of 400 is demonstrated at the 95% confidence level, unless the nickel-plated elements fail first. (b) Irradiate five columns of chemically nickel-plated C-64 clad OIIN fuel elements, alternating 0.2 mil and 0.5 mil plate, heat treated at 300{degrees}C for six hours. Two columns will be discharged at 400 MWD/T and three columns at 800 MWD/T exposure. (c) Irradiate approximately thirty columns of nickel-plated, C-64 alloy clad fuel elements …

Investigations of materials for use in connection with the SNAP I mercury Rankine cycle power conversion system are discussed. Test programs are outlined and results are tabulated for each candidate material. Several nonmetallic materials and processing procedures were developed which enabled uncooled high-performance electric machinery to operate at 550 deg F in mercury vapor. (J.R.D.)

The investigation of heat transfer in oscillating flow was continued. A series of tests to further investigate the dependence of heat transfer on various parameters was made for three locations of the heat-transfer section along the duct at several harmonic frequencies with pressure amplitudes varying from 0.4 steady-state pressure to zero. (W.L.H.)