In order to retain the hermetic feature of the sodium flow controller, magnetic flux linkage of a permanent magnet coupling is used to transmit the torque produced by the operator through the pressure wall. Magnetic calculations, manufacture, and testing of this magnet coupling are described. (M.C.G.)

Possible locations of the projected 40-tube overbore facility at the C Reactor are discussed from the standpoint of obtaining conversion ratio data applicable to a full-reactor overbore program.

System curves for each portion of the C Area Water Plant were obtained from referenced work and are presented in figures. Field test data, corroborating the calculated curves, are presented as singular points on the same graphs. Present maxima capacity of the C Area Filter Plant was 121,000 gpm with 118,000 gpm available for use as primary reactor coolant. Modifications to the filter effluent piping would increase this available flow to about 180,000 gpm. Of the 118,000 gpm available for C Reactor use, 10,000 to 12,000 gpm was demanded by B Area through the 183 BC intertie. The maximum flow that the intertie line could handle, without reducing the filter capacity of the C Area filters, is about 21,000 gpm.

105-KE downcomer pressure data are tabulated.

The object of this test is to determine the hydraulic characteristics of the proposed overbore process tube assembly designs which are to be installed on 105-C reactor for the 40 tube overbore fuel element test.

C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.

An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)

Descriptions are given of various procedures used in determining leaks in the tubes of the 1B heat exchanger. Air pressurization tests determined leakage and leak rate of nine tubes. The leak-location-detector-probe method was found promising for locating defects along the length of the tube. Results of the probalog, dye-penetrant, and ultrasonic tests proved inconclusive in determining leak locations. (B.O.G.)

The design and construction work completed on the 2000-kw Sodium Teat Facility during the period from April 1958 to Oct. 1959 is described. The purpose of the facility project in to test models of equipment components which are to be used in a moltsn plutonium, sodiumcooled fast reactor. (D. L.C.)

Alternative P-type dopants such as, Th, P, and As were studied. Ingots were grown from a melt containing one at.% dopant and their electrical properties evaluated. Also, sintered pellets of PbTe were doped by exposure at high temperature to gaseous dopants. In most cases, the doping concentrations obtained were insufficient for SNAP 10A requirements. In the case of As, however, doping of Te-rich PbTe, the desired heavy doping was obtained. These preliminary studies suggest that dopants other than Na might be suitable for SNAP 10A requirements. (auth)

It is necessary to provide accountability tables and equations for use in routine product buildup predictions for the duration of the E-N and blanket loads charged under PT-IP-350-C. Recommended values are given.

A survey was made of unclassified literature dealing with the control of activity build-up in pressurized water reactors by high-pH coolants. Survey results indicate that a high-pH coolant is desirable for control of induced radioactivity build-up in pressurized water reactors containing significant amounts of stainless steel and that no significant hazards are introduced by adoption of a high-pH coolant. Lithium hydroxide was found to be the most acceptable agent for coolant pH control. Strong-base-form mixed-bed resin is effective in maintaining both coolant pH and coolant purity. (auth)

The adsorptive capacities of various inorganic adsorbents and activated charcoals for krypton and xenon were determined. Columbia-G activated charcoal had the highest capacity for both krypton and xenon at pressures from 0.01 to 125 mm Hg and temperaturens from 2 to 85 deg C. If a value of 1 is assigned to the capacity of this charcoal at 28 deg C for krypton, other charcoals range from 0.63 to 0.84, molecular sieves (except 4A) from 0.11 to 0.20, and some silica genls from 0.05 to 0.07. Various othenr adsorbennts, including one variety of silica gel and molecular sieve 4A, range from 0.005 to 0.032. Molecular sienve 5A and Columbia-G charcoal adsorbed 11.5 times more xenon than krypton. Adsorption of 7.5% water by either of these adsorbents lowerend their capacity for krypton 25 to 30%, while saturating the sieve material ( approximates 15% H2O) lowered the krypton capacity 80%. (auth)

The purpose of this document is to assembly certain job elements which may become part of the Advanced Engineerng Subsection programs, so that those individuals who may be requested to participate in such programs may be aware of the nature and form of Advanced Engineering planning.

This is one of twenty-one volumes summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume summarizes the methods and techquies developed for use in the thermal design of nuclear reactors associated with that program.

Goats under Nembutal and mice were exposed in a shock tube modified to give longer-duration overpressures and equipped with ports for photography. Except for ear damage, displacement still seemed to be the major cause of injury in goats in the 4-foot section of the tube. However, it was found that mice were killed, apparently by pure blast, in a particular position in a particular type of side chamber attached to the higher-pressure, smaller-diameter, section of the tube. The relation between maximum velocity of displacement and the distance animals were thrown was fairly regular. The probability of injury over different terrains is discussed. (auth)

The use of oversized piping in an air lift for transferring solutions at rates less than 5 liters per hour was proven feasible with certain limitations. Reliable operation was also obtained with air lifts containing a lateral run inserted between the point of air injection and the final discharge point. Discharge of the air lifts, especially at low liquid flows, was very erratic under the conditions studied. (auth)

Studies the prototype and production air pulse units.

The limitations and operating techniques which were in effect at ANPD for the prevention of criticality accidents are summarized. The standards followed by the atomic industry were followed; however, the safe mass of U/sup 235/ moderated with beryllium oxide and hydrogeneous materials was based upon criticality experiments conducted at ANPD. Although the guide was primarily for the use of the ANPD nuclear safety control organization, it may also be of assistance to designers and operating management in maintaining nuclear safety. (auth)

This report includes a summary of materials research and engineering of the aircraft nuclear propulsion project.

Information is presented on a suitable amine processing system for plutonium nitrate. Experiments with concentrated plutonium nitrate solutions show that trilaurylamine (TLA) - xylene solvent systems did not form a second organic phase. Experiments are also reported with tri-noctylamine (TnOA)-xylene and TLA-Amsco - octyl alcohol. Two organic phases appear in both these systems at high plutonium nitrate concentrations. Data are tabulated from loading and stripping experiments. (J.R.D.)

One of the characteristics of the safety control system of the Hanford reactors which is of primary concern from the standpoint of reactor safety is the response time. The purpose of this study is to determine whether the safety control system is sufficiently fast-acting to prevent an intolerably large power excursion from developing. An intolerably large power excursion is one that would cause a melting of the uranium while the reactor power is above the previous operating level. The purpose of this report is to show the techniques used to analyze the speed of control of the Hanford reactors. This report will help to standardize the methods and reduce to a routine the solution of this type of problem. The analog circuits will be presented and explained. A report has been written previously on the speed of control. The changes in the problem have necessitated the redesigning of the analog circuit and also the issuing of this report. This report compiles and organizes the analog work done on the speed of control to date.

None

This document discusses a proposal that certain presently installed reactor process water outlet temperature data logging equipment in subject reactors to be replaced with new functionally simplified equipment of a more adequate design. The primary purpose of the proposed installation is to replace existing equipment which is obsolete and in three reactors is worn out to the point where the equipment is out of service frequently for periods of time up to 8 hours or more. The new equipment will provide reliable process tube temperature information for use in the functions of reactor control and product accountability. Based upon anticipated incremental production gains resulting from use of the new equipment, the amortization period for the project is calculated at 2.7 years.