Machining was completed on the s-tube flow element, and a spring, for use at a constant known temperature, was constructed for the flow element. Calculations for geometry factor, detector efficiency, and effective absorption along the useful radiation path were completed for the reed densitometer. The flow rate control was calibrated and performed satisfactorily. Response to step transient changes occurred with a damping constant of about 0.6 and an effective response time of 10 sec. Sources of error in the s-tube flow rate measurements are discussed. (M.C.G.)

The means used to prevent the recurrence of tetralin leakage into the SRE sodium systems are discussed. Included is a description of the redssign of system components to utilize alteraate coolants such as nitrogen, air, and NaK. The use of kerosene to replace tetralin, where double containment is provided, is discussed. The physical properties are compared, and kerosene is shown to be free of the undesirable characteristics of tetralin. The fuelelement cleaning systsm was redesigned for steam washing, followed by a water rinse and vacuum drying. Hydrogen gas evolved during washing is oxidized with copper oxide to eliminate the possibility of a hydrogen-oxygen explosion if air should accidentally enter the vent system. The fuel element was changed from a seven- to a five-rod cluster to provide additional clearance in the channel. Element hardware was modified to provide an orifice location which will permit more precise flow calculations; a redesigned hanger assembly which will minimize sodium holdup; and filter screene at the channel entrance to prevent in-cluster plugging. Diagnostic instrumentation was provided for the reactor to monitor: fission-product activity in the reactor cover gas; rapidly varying reactor parameters during a reactor scram; internnl and aheath fuel temperatures; and fuinctions being performed …

A theoretical treatment is made of the stress-strain and plastic flow properties of spherical shells of stainless steel under dynamic loading. The results are used to calculate the response of a homogeneous reactor core to inential pressure pulses. The importance of the duration of the pressure pulse as compared with a natural time or period of the material in plastic as well as in elastic deformation is noted. For a small homogeneous reactor core vessel, the natural plastic "swing time" is found to be about 0.25 msec. The rise time of the pressure pulse is 2.5 msec while the decay time is approximately 12.5 msec. Thus the rise time of the pressure pulse is about 10 times as large as the ""swing time,'' which means the static pressure approximations hold quite well, and maximum energy absorption is determined for the conditions under which the vessel will fail. (auth)

This document details activities of the irradiation processing department during the month of January, 1961. 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; Financial Operation; and NPR Project.

This is the monthly report for the Hanford Laboratories Operation, January 1961. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

The objectives of this test are: 1. To establish grain size limits for acceptable uranium fuel element cores. 2. To establish, if possible, criteria for predicting core dimensional stability during irradiation by comparing the relative dimensional stabilities associated with grain size and with variations in grain size in individual cores. 3. To obtain process tube and fuel corrosion data associated with bumper fuel elements in new tubes with no mixer, one mixer in the 10th position and two mixers in the 7th and 15th position from the rear. Fuel cores representing the full range of UT-2 voltage values (grain size converts to d-c voltage) of interest are segregated into three categories: a. Large grains. b. Variations of grain size in an individual core. c. Small grains. Each category will be subdivided into three groups, each covering a small range of values. After canning, the finished fuel elements will be assembled into twenty-seven (27) charges in three latin square patterns for irradiation to a 900 MWD/T exposure goal in D Reactor.

This report discusses benefits in an increase in the flow of water from the lower depths of the Grand Coulee Dam which was used to lower river temperature at HAPO. A net average daily reduction of over 1.2{degree}C resulted at HAPO with a peak of 2.7{degree}C. The Net Production gain from temperature change was 6910 MWD and the Cost of Control was: Grand Coulee Charges $3,120.00, and other (Estimated) 6,880.00 for a total of $10,00.00.

The flux of hydrogen gas through Hastelloy B, hot forged to 20% reduction was determined as a function of membrane thickness, pressure differential, and temperature. The experimental data were fitted to the equation J = (1810 P/sup 1/2//X exp (-8422/T) where J is the flux (cc at STP/ (cm/sup 2/- hr)); P is the pressure (atmospheres); X is the membrane thickness (mm); and T is the temperature ( deg K). (auth)

A hydraulic analysis of the Pathfinder boiler core was completed. A description of flow paths including a schematic diagram is included. Pressure drops along primary and leakage fiow paths were calculatsd. After the resistance of each of the leakage paths was identified and the available driving forces in terms of pressure drops associated with the active fiow were established, the leakage rates were easily determined. The dependence of pressure drops on core operating parameters was investigated. The boiler flow rates were also determined. The natural circulation flow rate as a function of power level was estimated at normal condltions of temperature and pressure. The relationships between the boiler parameters of fiow, power, pressure, and inlet subcooling which arise through the laws of conservation of mass and energy are presented. (M.C.G.)

A calculation of the three-group flux distribution of the initial wet critical loading of KW reactor was made using the same calculation techniques as those employed in NPR three-group calculations. The calculation results are directly comparable to experimental data obtained during the KW loading. The calculated radial geometrical buckling of the critical loading is 48.2 {mu}b compared to a measured value of 49.8 {mu}b; these agree within estimated experimental error. The calculated reflector augmentation distance is 54.5 cm compared with a measured value of 49.2 cm. The three-group nuclear constants of the KW lattice components were calculated from 18-group data (APEX-515) based on the flux spectrum of HW-63172. A cell calculation was made to determine the effective constants for the homogenized cell. The calculated k {infinity} of this cell came out 0.998 compared to a measured 1.034. Since the resonance capture of U-238 is difficult to treat, in three-group theory, all error was assumed to be in this parameter which was adjusted accordingly to give k {infinity} = 1.034. The resulting core parameters are given in Table I, along with parameters for the dry, unloaded core and the wet unloaded core.

The evaluation of Kanigen electroless nickel plating for surfaces in contact with water and steam in a sodium-heated Type 316 stainless steel steam generator is reported. The purpose of the coating is to afford protection from stress corrosion cracking originating on the water-steam side of the unit. It is concluded that the Kanigen coating does not afford adequate protection for the service conditions. (D.L.C.)

Modifications of EBWR are described, and data from analysis of control- rod operational crud deposit are tabulated. Development and construction progress on Borax V are summarized. Research for ZPR III is reported on a series of uranium oxide critical cores. Comparative data on composition, critical masses, and central fission ratios are tabulated for these assemblies. Developmental work on ZPR-VI, ZPR-IX, and Juggernaut is summarized, followed by a summary of EBR-I operation and development of EBR-II. Studies of reactor steam superheating concepts are reported from which conclusions are listed. In reactor safety studies, data were obtained on the extent of water reaction with various metal combinations for use in cores as function of energy input. In other work in-pile melt down studies were conducted. Operation and development of TREAT are described. Nuclear technological developments are summarized in sections on physics, fuel development, reactor components and materials, separations development, and advanced reactor development. (For preceding period see ANL- 6295.) (J.R.D.)