ABS> The removal rates of iron, nickel, and chromium from synthetic stainless steel waste solutions during electrolysis over a mercury cathode were studied. The loading capacity of the mercury for the stainless steel metals was estimated on the basis of laboratory experiments to be about two% by weight. The laboratory data indicated that, at an electrode potential of --1.80 voits vs S.C.E., 85 ampere-hours per liter of waste removed essentially all of the stainless steel ions from a sulfuric acid solution containing 0.13M metal ions at 35 deg C. (auth)

Progress is reported on thermonuclear research. Separate abstracts were prepared for 8 of the 10 sections. Design and engineering service reports and notes are given in the remaining sections. (M.C.G.)

This report attempts to present some of the factors which must be examined and the issues which must be decided in considering a proposal for national welfare standard, as well as a summary of the arguments which can be made for and against the adoption of national standard.

This report provides a bibliography of resources related to the seniority system in Congress.

Using the group vibratibn method of McMurry, the normal frequencies and coordinates of propane and three of its symmetrical deuterium substituted compounds were obtained. The force constants used were taken from a variety of previous works on hydrocarbons. The results give reasonable agreements with the experimental frequency and mode assignments of others. (auth)

A test was conducted on August 12 to 13, l961, on the operation of the four reactor coolant loop relief valves and the four purification system relief valves. The results indicated proper operation. (D.L.C.)

The multiple hydroclone system removed l83 grams of corrosion product solids in l814 hours of operation during reactor run 2l. The low removal rate was attributed to plugging of multiclone feed ports that presumably occurred during the latter part of run 20. After modifications to the reactor core and removal of the multiclone unit at the end of run 21, the reactor core was backflushed with the flow direction in the core loop reversed. During this period, the single hydroclone removed 205 grams of solids in 10.5 hours of operation. (auth)

The present method used for handling multichannelanalyzer data at the ORNL 86-Inch Cyclotron is stated. FORTRAN subroutines for reading the analyzer data into the IBM-7090 computer and for printing out the processed data and punching processed data on cards are presented. (auth)

The SNAP-2 Primary System Test loop fabrication was completed with associated flight prototype components including reactor core and boiler mockups for volume and DELTA P simulation, CRU-IIII NaK pump, compact heater, and expansion compensator. A mobile loading system was designed and fabricated with the capability of cleaning the NaK prior to final loop sealing. Loop descriptions, test objectives, and operating procedures are presented. (auth)

A discussion is presented concerning the preliminary design and hazards evaluation of a H-cooled in-pile experimental loop for operation in the large double-walled through-tube in the Experimental Gas-Cooled Reactor (EGCR) at Oak Ridge. This loop is designed to permit experimentation with full-scale fuel element configurations up to 8 in. OD, at inlet gas temperatures of 600 to 950 deg F at 300 psig, and experimental power levels up to 500 kw. The results of a preliminary hazards evaluation indicate that a loop of this type can be safely operated in the EGCR. The H flammability hazard is controlled by blanketing all H-filled pipes and components with a sufficient quantity of nonreactive gas, such as He or CO/ sup 2/, to produce a noncombustible mixture for all credible H- release situations. (auth)

Thermal stresses in the HFIR beryllium reflector were computed for the unlikely case where the reactor is scrammed with a simultaneous loss of coolant flow and for the case following an electrical power outage where the reactor power level and the coolant flow rate are reduced simultaneously. For the case where the reactor is scrammed with a sudden loss of the coolant flow, the resulting maximum tensile thermal stress following the scram is 22,500 psi. In case of an electrical power outage, the maximum tensile thermal stress following a reduction of the fission power level from 100 Mw to 10 Mw with the lowering of the coolant flow rate to 10% of the normal value is 12,800 psi. (auth)

Since brazing was the method selected for joining the stainless steel SM- l reactor fuel element, corrosion studies were conducted on various potential brazing alloys to evaluate their resistance under the approximate pressurized- water conditions of the SM-1. The program consisted mainly of testing type 304L stainless steel T'' joints brazed with selected alloys in quiescent, degassed, and deionized autoclaved water at 565 deg F under 1200-psi pressure. In the initial phase of the investigation, tests were limited in duration to l000 hr in order to quickly screen some 18 potential alloys for longer time testing. Based on weight-change data and the metallographic examinations, five of the 18 alloys exhibited sufficient corrosion resistance to warrant further investigation. These alloys were subjected to autoclave tests of 12 and 16 months. In these extended tests, 1 cc O/sub 2/liter and a mixture of 1 cc O/sub 2/liter plus 50 cc H/sub 2/liter, respectively, were added to the water to more closely simulate SM- 1 reactor water conditions and to evaluate the effect of different gaseous additions on the corrosion behavior of the alloys. On the basis of weight-change data and metallographic examination after long-term exposure of the tested stainless steel-base joint; these …

Investigation conducted to determine the cause of cracking, during tension, on the surface of tensile specimens of uranium-molybdenum alloys.

In order to study resonances in the {Sigma}-{pi} system, we have analyzed reactions in which a {Sigma} hyperon and two or three pions are produced in K{sup -}-p interactions at 1.22 {+-} 0.040 and 1.51 {+-} 0.050 GeV/c incident K{sup -} momentum (i. e., 1895 and 2025 MeV center-of-mass energy), using the Lawrence Radiation Laboratory's 72-in. hydrogen bubble chamber.

From abstract: This report presents information generated during a six month feasibility study of an engine which uses a supercritical working fluid as the secondary portion of nuclear powered electric generating system.

Quarterly report describing the progress and development of improved zirconium alloys for service in superheated water and steam. This report covers the period between October 1 to December 31, 1961 and was conducted by the United States and the European Atomic Energy Community (EURATOM).

Quarterly report describing the progress and development of improved zirconium alloys for service in superheated water and steam. This report covers the period between July 1 to September 30, 1962 and was conducted by the United States and the European Atomic Energy Community (EURATOM).

A thermal analysis was carried out to determine the temperature distribution in a SNAP-8 type reactor system during atmopsheric reentry. Of particular interest are the temperature distributions in the reactor upper head, the upper grid plate, and the vessel wall. The time and altitude were determined at which the upper head falls away from the reactor core due to having a portion of its wall melted through. The time and altitude of the melting of the upper grid plate and vessel wall were also determined. The effects of reentry attitude or equivalent angle of attack and initial temperature on the thermal behavior of the system were investigated. The computer programs used in various phases of the analysis were NEWTON (drag coefficient), RESTORE (reentry trajectory), and TAP (thermal model).

This document summarizes the amounts of radioactive contamination discharged to ground from separations facilities through December 1962. Detailed data for individual disposal sites are presented on a month-to-month basis for the period of January through December 1962. Previous publications of this series are listed in the bibliography and may be referred to for specific information on measurements and radioactivity totals prior to December 1962. Tables list the major disposal sites in the separation facilities, total volume of waste discharged to each location, and the gross amounts of plutonium and beta particle emitters discharged to ground since startup. This same data is presented on a monthly basis for cribs still in use. Information is presented on the source of the waste stream and the settling facility if used. Isotopic data are included for disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to swamps are also included.

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The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

This document details the monthly activities of the Reactor Branch for the period of January 1960 through December 1960.

A procedure is described for the sequential separation and determination of Arsenic-76 and Phosphorous-32 in reactor effluent water. The analysis can be performed in less than one hour and the arsenic fraction is scanned immediately after separation on a multi-channel gamma spectrometer. The discrete gamma energy of .56 MEV for Arsenic-76 makes further purification unnecessary. The organic phase containing the Phosphorous-32 is dry mounted on a one-inch stainless steel dish for beta counting.

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