The types of tests performed in SPERT-I, and the tests proposed for SPERT-II and -III, are described. These reactors are described, and factors influencing their dynamic behavior are discussed. The tests are classed as static, step, ramp, and oscillatory. The correlation between the test results and the reactor dynamic safety characteristics (stability, self-shutdown under excursion conditions, etc.) is investigated. (T.F.H.)

The iron thermal shields of the Hanford reactors are cooled by means of water flow through thermal cooling tubes embedded in the shield blocks. The flow rate, temperature rise, and allowable pressure in the tubes and the conditions under which some of the tubes may be out of service are specified in the process standards. Shield heat formation and heat transfer calculations are necessarily based on broad assumptions and are therefore usually reliable only for order of magnitude and trend predictions. In order to specify condition of shield cooling under which the reactors may be safely and economically operated data must be available regarding operating temperature as related to the flow of cooling water through the shield. Some of the data on which the current standards are based have been determined by extrapolation to present conditions of measurements taken several years ago. The purpose of this test was to establish current data that may be used in updating the thermal shield coolant standards. Measurements were taken of the operating temperatures experienced in an uncooled thermal shield cooling tube in relation to the specific power of the adjacent process tube. Conditions were varied by adjusting the flow in the thermal shield …

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A new outlet elbow fitting has been designed for use on C reactor to replace those elbows which fail because of stripped threads where the elbow connects to the rear nozzle barrel. There are two models of this threadless elbow, one providing for a water thermocouple only, and one providing for a thermocouple and an RTD. Flow tests have been conducted to determine the pressure drop across these outlets and the pressure drop has been compared to the pressure drop across the previously used outlet elbow and pigtail assembly. Results of these tests and comparisons are shown.

The outlet water temperature limits for the Hanford production reactors are based, among other things, on the ability to prevent excessive fuel temperatures during or following inadvertent flow losses or power surges. The analyses involved in the determination of the limits are dependent upon laboratory heat transfer experiments which closely simulate actual reactor conditions in a single process tube. The heat transfer experiments are quite complex in that they involve the recording of rapid changes in pressure, temperature, flow, and heat generation rates. As a preliminary to these transient heat transfer experiments, selected steady state experiments are performed to better plan the experimental program and to aid in the analysis of the problem. This document presents the results of some steady state experiments performed at low tube powers for K-reactor. The results are reported separately from the transient experiments because the data are useful for reactor problems involving shutdown cooling and are also of value to people interested in pressure drop of two phase mixtures.

This report, for June 1962 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and safety and security.

This report details activities of the Metallurgy Development Operation for the months of April, May, June 1964.

Tubular thermoelectric module technology was advanced during Phase II-C of the Compact Converter Program to the extent that the baseline temperature for module testing was raised from 1000°F 1100°F. Significant development work was performed to produce a system utilizing a heat pipe that transmitted thermal energy from an external heat source to a tubular module.

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This report addresses the buckling tests of scaled graphite barrel inner reflectors.

A spectrographic method was developed for the complete cation analysis of high-purity water by spark excitation of the evaporated sample on pairs of silver electrodes. The procedure is fast and precise, with a minimum of sample handling and adequate sensitivity for determining verv small amounts of trace impurities. (auth)

A detailed analysis of the PM-2A steam generator tube sheet revealed that is is safe from strain cycling damage. The pressure stresses, however, indicated that permanent deformation would take place in the tube sheet at hydrostatic test pressure of 1.5 times design pressure. (auth)

Compressors were required to circulate helium coolant in an in-pile loop at moderately high heads (6000--8000 ft), at low volumetric flows (30--40 acfm), at 600 deg F and 385 psig. A regenerative type of gas compressor, using greaselubricated ball bearings enclosed in a special water-cooled casing, was designed and tested. The design objectives were achieved, and a number of these compressors have been built and are in operation. (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)

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