This report, from the Chemical Processing Department at HAPO for January 1961, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

A relationship is derived between the mean square fluctuating current of a hot wire anemometer and the six turbulent Reynolds' stresses in the stream- coordinate system without employing the usual low turbulent intensity approximation. The relatively simple result is a consequence of assuming proportionality between the wire current reading and the perpendicular velocity component instead of the nonlinear dependence required by King's law. The assumption is valid for instruments equipped with the proper linearizing circuitry. Tbe stream-coordinate Reynolds' stresses are then related to the cylindrical polar Reynolds' stresses. An error analysis on the experimental determination of ore of these stresses is indicated but cannot be evaluated without fu rther data. (auth)

This report, from the Chemical Processing Department at HAPO for February 1961, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

The transport of fission products from points of origin in unclad graphite matrix-type fuels to the reactor circulating system involves, as one of the steps. diffusion through the graphite matrix to the fuel element surface. As pointed out by Rosenthal. the fraction of a given fission product chain actaally reaching the fuel element surface will be small if the time for transport through the graphite is long compared to the half-lives of the volatile members. An important problem, therefore, is the determination of the effective transport rates of the various mobile elements and their daughter products of interest through various graphites suitable for use as fuel element compacts. as functions of temperature over the range of greatest immediate interest to reactor designers. The upper end of the range need not exceed about 1000 deg C. The transport of helium and arbon through various graphites has been the subject of considerable study by Watson. Evans, and other,. and a prelimilnary investigation of the high temperature transport of some ordinarily non-volatile elements has been carried out by Saunders. This work is briefly reviewed in relation to the final problem and the areas in which further information is needed most by reactor designers …

This report for March 1961, 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; and employee relations.

Three Pu preparations were made and partially characterized. These preparations were one pure PuO/sub 2/ preparation via the batch-oxalate route and two PuO/sub 2/-UO/sub 2/ preparations containing 0.5 and 20 at.% PuO/sub 2/ via the continuous coprecipitation route. The PuO/sub 2/ preparation, which had the largest surface area of the preparations (29 m/sup 2//g), tended to pick up weight when exposed to the box atmosphere. Dry pressing and sintering studies were conducted on the PuO/sub 2/ preparation. The results indicate that the sintered density of PuO/sub 2/ pellets is independent of green density above 5.6 and 6.0 g/cm/sup 3/ for pellets sintered at 1600 and 1400 deg C, respectively. Preliminary reactor physics calculations for near-thermal reactors indicate that extremely long core lives can be attained with a fuel of initial composition 78 at.% Pu/sup 239/, 12 at.% Pu/sup 240/, and 10 at.% Pu/sup 24 / 1>s, and that the reactivity variation during core life is nominal. (D.L.C.)

Production of Pu and U nitrates, U oxides, Pu metal buttons, and fabricated parts are reported. 135 kCi Sr-90 was isoalted in Purex plant and stored for further Sr-89 decay. The new scrub cartridge in Purex co-decontamination cycle column performed better than the old. Increased Np losses were attributed to failure of plastic plates in 2nd cycle U extraction column.

This report addresses the configuration of Tatum Salt Dome, Lamar County, Mississippi.

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This report, for June 1961 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; weapons manufacturing operation; and safety and security.

The effectiveness of purification in controlling plant radioactivation rates was determined through comparison of the actual effects produced from plant operation with and without purification. With the purification system in service, there was no appreciable difficulty in maintaining the reactor coolant within reference water specifications. In addition, there was no discernible increase of crud deposition in the coolant system as determined by direct radiation measurements of the purification hairpin loop. However, without demineralization as a controlling agent, the gross non-volatile gamma activity levels of the reactor coolant increased and the specific activities of longlived fission products were at higher levels. Although the general levels of water- borne activities increased during the test period, the associated plant systems did not exceed their limits. (auth)

This report, for July 1961 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; weapons manufacturing operation; and safety and security.

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On the 0000 -- 0800 shift, December 9, 1960, the delayed neutron monitor on KER Loop 3 gave a high coolant activity signal indicating a possible fuel element failure in this loop. KE Reactor was shut down immediately thereafter. This report is being written to summarize the events pertinent to this KE Reactor scram and to discuss the results and significance of data from analyses on coolant and coupon samples taken from the KER Loop 3 system.

This report, for August 1961 from the Chemical Processing Department at HAPO, discusses the following; Production operation; Purex operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

Activities in research programs are summarized in the areas of power reactor fuel processing, fluoride volatility processing, molten salt reactor fuel processing, homogeneous reactor fuel processing, waste treatment and disposal pilot plant decontamination, GCR coolant purification studies, equipment decontamination, HRP thoria blanket development, fuel cycle development, transuranium element studies, production of U/sup 232/, uranium processing, fission product recovery, thorium recovery from granite, solvent extraction technology, mechanisms of separation processes, radiation effects on catalysts, ion exchange technology, chemical engineering research, chemical applications of nuclear explosions, reactor evaluation studies, and assistance programs. (J.R.D.)

A development program was initiated under Project CG-775 to test the proposed. process water pump modifications for Project CGI-883, Increased Process Water flow 100-K Area. Both the high lift and low lift pumps of the 190-KE pumping Unit were modified during May, 1959 for in-service prototype testing of the proposed equipment revisions. The testing proved the feasibility of the modification program but a problem, in the form of capitation damage to the high lift pump impeller, was noted. An incremental production gain at the NE reactor was realized as a result of the test program. Interim funds were authorized in February 1960 to initiate detail design on total KE and KW water plant expansion and to modify one KW process paying set for an early incremental production gain at the KW reactor. The modified KW process pumping unit was placed in service in mid-1960 and a proportionate reactor production gain was realized. Shortly after the modified KW pumping unit was placed in service it was noted that the total process water flow in KW was somewhat greater than in KE. An investigation was made to evaluate the low lift pumping Units in each area to find out if they contributed …

The history of gas-cooled reactor developmert is reviewed. The 40-Mw HTGR r the Peach Bottom Power Plart is described. Other projects to follow this prototype are outlined. Fuel development and progress in graphite core technology are discussed. The program sponsored by the Empire State Atomic Development Associates for a high-temperature gas-cooled reactor with an output of 300 to 500 Mw(e) is described. (M.C.G.)

Theoretical and experimental studies indicated that the effective permeability coefficient for graphite is lowered by a helium stream in-sweeping through the graphite pores. This phenomenon was considered in the design of HTGR fuel elements. A portion of the helium gas which is drawn into each fuel element as a purge stream may enter through porous wall sections, supplementing the purge gas entering at the top of each fuel element. The purge stream leaves each fuel element through a header system which carries the purge gas to an external fission product trap. The flow rate through the trapping system determines the upper limit of the average in-leakage through the fuel element walls. In the case of the HTGR, a graphite having a helium permeability of 1.1 cm/sup 2//sec at 350 psia and 700 deg F (approximately 0.1 cm/sup 2//sec at 14.7 psia, 70 deg F) would result in 100% of the purge flow entering through the wall sections of the fuel element. A lower permeability graphite, with most of the purge flow entering at the top of the fuel element appeared to be more desirable for maintaining optimum purge flow conditions. (auth)

It is noted that a uniform flat-top beam spill-out cannot be obtained in the Cosmotron by manual control. A servo system is proposed which will control this spill-out by sensing the external beam intensity, and correcting the magnet voltage to keep this intensity constant. This servo must operate through the transfer function of the main ignitron system and the flat-top filter. An analysis of these special transfer functions is presented. (J.R.D.)

This report for November 1961, 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 special separation processing and auxiliaries operation.

Ring-joint flanges were studied in thermal-cycle tests as part of the development work associated with Homogeneous Reactor Experiment No. 2 (HRE-2). The purpose of this study was to provide criteria for design, installation, and operation of joints that would remain leaktight under reactor operating temperatures and pressures. Joints ranging from 1/2 in., l500 lb to 4 in., 2500 lb and with various initial bolt loadings were cycled between room temperature and 636 deg F. It was demonstrated that when joints were made up to HRE-2 standards and specifications, leak rates of less than 0.25 x 10/sup -3/ g of water per day per inch of gasket pitch diameter could be routinely- attained. Undamaged gaskets could be reinstalled or new gaskets used with equal probability of achieving acceptable leak rates. The system installed in HRE-2 was provided with a high-pressure buffer system to ensure that the small amount of leakage to the cell would be nonradioactive. (auth)

Development activities on the large-scale, hightemperature, thoria slurry facility at ORNL were suspended in July 1960. The latest design philosophy, the status of component development, operational procedures, and limitations of each section of the system are presented. The design of the facility would be adequate for a singleregion slurry-reactor experiment on a slurry blanket system with a properly developed blanket flow pattern. The equipment consists of a 330-liter circulating system, including a 32-in.-ID pressure vessel designed for service at 2000 psi and 300 deg C, a 3000-lb- capacity low-pressure system, and the interconnecting lines. Flow sheets are included and materials of construction and recommended instrumentation are discussed. The first five experimental runs in the system, which were made to determine the feasibility of operating with fertile thoria slurry in the blanket region of Homogeneous Reactor Experiment No. 2, either as-built or after minor modification of the original system are summarized. The first run, a heat exchanger test run made at 1800 psi and 300 deg C, was considered inconclusive. Repeated plugging was experienced, but this occurred in the connecting lines rather than in the heat exchanger. The next four runs were investigations of the high-pressure blanket system, with the pressure …

Historical developments of the ORNL dosimeter and the ORNL badge meter, Model II, are discussed. Beta-gamma and neutron dosimetry techniques are described. ORNL film dosimetry experiments are summarized. (M.C.G.)