For presentation at ASCE Convention in Reno, Nevada on Thursday, June 23, 1860. A description of the Hallam Nuclear Power Facslity is presented. The history of the project, program participants, site description, component development program, reaetor building, reactor structure, reactor core, sodium systems, instrumentation and control, fuel and component handling, auxsilary sustems, special design features, and advantages of sodium graphite reactor systems are discussed. (M.C.G.)

None

Development of a radioisotope-fueled thermoelectric power-conversion system for the SNAP program is discussed. The generator operates at somewhat less than 1% over-all efficiency, has a power output of approximately 1.45 watts, and has a specific power of 0.l42 watts per pound. The power-flattening device did not perform as expected. A second generator is also described which produces a maximum power of 1.6 watts, has a maximum efficiency of 1.20%, and has a specific power of 0.32 watts per pound (exclusive of the weight of the heat source). (auth)

An apparatus for determining linear thermal expansion data up to 1350 deg C for materials that cannot be heated in air is described. Expansion is not measured directly, but is derived from the differential expansion between the material being tested and the materials of construction of the dilatometer. The apparatus is calibrated against published data on the linear thermal expansion of both fused silica and polycrystalline tungsten metal. It can be operated either as a high-vacuum or as a controlled-atmosphere apparatus. Recording of data is either fully automatic or manual. The percentage error is believed to be no greater than plus or minus 0.1 to 0.2%. Original data for linear thermal expansion of materials such as hot-pressed BeO, hafnium-free ZrC/sub 2/, zirconium-free HfO/sub 2/, INOR-8 alloy, siliconized silicon carbide, uranium dioxide, compositions of Be + BeO, and samples of CS 312 graphite, boron nitride, and tungsten metal. (auth)

None

A study has been made of the wear and friction phenomena associated with sliding molybdenum surfaces lubricated by liquid sodium. Operating conditions included sodium temperatures from 250 to 1300 deg F, sliding speeds of from 0.22 to 2.5 nan per sec, and contact stresses of 80,000 psi. Sliding specimens consisted of a ball and a flat rectangle. Sodium was found to influence the sliding behavior of molybdenum by modifying reacted films which provided boundary lubrication. These effects were also found to be temperature dependent. Addition of clean sodium between dry rubbing molybdenum surfaces in an argon atmosphere of high purity produced a marked drop in friction coefficient, and the character of the wear scar changed from galling to superficial scratching. Furthermore, these effects persisted after removal of sodium by evaperation. Studies of the films formed on molybdenum surfaces during sliding in liquid sodium and under controlled atmospheres were carried out using electron and x-ray diffraction, electron microscope, and differential thermal-analysis techniques. It was established that a sodium molybdate film on the molybdenum surfaces can be associated with low friction and diminished surface damage. Studies in vacuums up to 10/sup -6/ mm of mercury showed that native oxides on molybdenum also …

None

None

None

This report presents a brief survey of the current status of technology on organic-cooled reactors.

The current technical and economic status for each reactor concept in the Civilian Power Reactor Program is summarized. The individual techical status reports which present detailed information will be published by AEC as TID-8518. Included in this summary are: power costs for various reactor types versus coal- fired plants; construction schedule for heavy-water natural-U reactors; and annual program costs. (T.R.H.)

The status of technology of nuclear power reactors in 1959 is reviewed. General research and engineering development activities are discussed. The reactors considered include the pressurized water, boiling water, light water moderated superheat, organic cooled, sodium graphite, gas cooled enriched fuel, gas cooled natural uranium, fast breeder, aqueous homogeneous, and heavy water. Power costs are compared with the cost of power from conventional plants. (C.H.)

The status of pressurized water reactors on a technical, economic, and operating experience basis is reported. Items covered include concept description, PWR objectives, general research and development prcgram, reactor data, reactor illustrations, operating experience, construction and operation schedules, and present limitations and problems. (W.D.M.)

The status of a program to develop a thermal breeder power reactor which is based on the Th-U/sup 233/ cycle is discussed. The general characteristics of two-region power breeder reactors which are either solution-core slurryblanket or slurry-core slurry-blanket are described. A review of studies on corrosion, physical properties, and metallurgy of fuel materials, in particular aqueous ThO/ sub 2/ suspensions, is presented. Development studies on cladding alloys are discussed. The processing of fuel and blanket materials for removal of fission and corrosion products and recovery of fissionable material is discussed. A flowsheet for the preparation of ThO/sub 2/ slurries is contained. Development studies on reactor components, systems, instrumentation, and control are reviewed. Reactor-control, safety, and criticality studies are discusspd. (C.J.G.)

Estimates indicate that, within the D/sub 2/O-moderated natural-uranium- fuel reactor category, the lowest cost reactor will be cooled by boiling D/sub 2/ O. The margin over nonboiling is small, especially for reactors in the 100Mw(e) size and smaller. A major area in which research must be advanced before any of the reactor concepts can be successful is fuel-element fabrication and irradiation behavior. Both oxide and metallic elements have their own advantages, and each type should be developed. Other areas in which additional work must be done are the reactivities of various fuel lattices. fuel element burnout (especially for the boiling regime), zirconium pressure tubes and their connections to steel, and methods of reducing and detecting D/sub 2/O leakage. (auth)

This report addresses the status report on boiling water reactor technology as of 1959.

The current development status of the sodium graphite reactor concept is described. The development history is summarized, and all important areas of development are discussed. The discussion of the SGR program is broken into three categories: (1) general research and development, dealing with reactor physics, fuels and materials, components, etc., (2) experimental reactors, operating or in some phase of design or constnuction, and (3) power demonstration reactors, operating or in a design or constnuction phase. (W.D.M.)

A description of fast reactor types is given and the objectives of the AEC program on fast reactors are outlined. General research and development programs completed and in process on physics, fuel and materials, heat transfer, fluid flow, coolant chemistry, reactor safety, and components and systems are discussed. Reactors completed, under design and construction, and under study are described. (W.D.M.) A general description of ship and power plant, tabulation of principle characteristics, economics, potential advances, and necessary research and development are discussed in detail. (W.D.M.)

The technology of natural-uranium-fueled graphitemoderated gas-cooled reactor power plants is summarized for its relevance to the technology of enriched-fuel graphite-moderated systems. The technology of D/sub 2/Omoderated gas-cooled reactors is also summarized. Estimated technical performance parameters are given for the enriched-fuel prototype and for a large natural- uraniumfueled plant. Current technical status is discussed in terms of reactor physics, heat transfer and fluid flow, core materials, components, plant design and conctruction, and hazards. Detailed tables of characteristics for various reactors are given. An extensive bibliography is included. (W.D.M.)

In developing the program plans presented, primary attention was given to the technological development required to achieve economically competitive nuclear power. The program objectives are given in brief and the specific reactor program is discussed in some detail. The nuclear technology program is summarized. (W.D.M.)

In an attempt to prove the feasibility of liquid sodium as a reactor coolant, a design stady of three vital system components was undertaken: the intermediate exchanger, the boiler, and the superheater. Investigation shows that, during transient changes in load, the steam and sodium systems' temperature changes will not significantly exceed the small steady-state temperature changes, provided that the control system mairtains the design steady-state ratio between the flow rates and power level during the transient. An analog computer stady of the controlled system behavior for normal and casualty control is discussed. (W.D.M.)

None

This report is the first quarterly report under Contract AT(30-1)-2528 for the decontamination and recovery of precious metals from contaminated scrap. This program encompasses the recovery of gold, platinum, iridium, rhodium and palladium from scrap and contaminated laboratory ware, equipment, etc. which may be contaminated by alpha emitters (i.e., uranium, plutonium and possibly polonium or combinations of these), beta-gamma emitters such as fission products or all three sources of radiation. In addition the scrap can be chemically contaminated by base metals, graphite or other contaminants. Osmium, ruthenium or silver are not to be recovered. Since the scrap can exist in such a great variety of forms such as platinum-rhodium laboratory crucibles, graphite crucibles containing platinum-base metal alloys derived from vacuum fusion baths, insulated thermocouple wires, reactor liners, porous platinum filters, alloys such as Baker alloy 413 (60 Au, 25 Pd, 15 Pt) it may be necessary to devise specific methods for a given type of scrap. Nevertheless the basic chemistry is the same and information and methods applicable to single systems will apply directly to the more complicated ones. Accordingly the relatively simple systems would be investigated first and the more complicated ones later in the program. An essential and …

Since the stress analysis of a complicated statically indeterminate structure involves the handling of many simultaneous equations, the S0207 code for the IBM-650 was developed at Bettis to solve this problem by the use of matrix transformations. The code has two parts, each with a title: (a) the CTAC code gives stresses, deflections, and reactions: and (b) the MODE code calculates natural frequencies and modes of vibration. As an aid to users of the code, a review is given of the elasticity and matrix transformation theory utilized in this code. (auth)