Sodium Technology for Nuclear Power Plants. F. A. Smith. 30 p. $0.50 (OTS). Sodium-cooled power reactors will require sodium flow rates through the reactor of 10,000 to 50,000 gpm at pressures up to 100 psi and temperatures up to 1,000° F. The present sodium-cooled reactors are smaller than required for central power plants; however, a program including development of pumps and piping systems, sodium pre-heating methods, instrumentation, heat exchangers, mechanisms and seals, and sodium chemistry has been instituted. A large composite test facility duplicating the salient features of the EBR-II primary sodium reactor system and containing 5,000 gallons of sodium has been in operation since 1955. Further work has been done on sodium purification for higher temperature systems, permitting the use of oxygen sensitive materials. A broad program of investigation has been initiated in the field of heat exchanger design, and several large sodium cooled reactors are under construction which will contribute extensively to the technology of sodium. (A/CONF. 15/P/2291-Argonne National Lab., Lemont, Ill.) Aqueous Homogeneous Reactor Fuel Technology. 53 p. $0.50 (OTS), Development of the technology of uranyl sulfate solutions and thoria slurries for power breeder reactors is outlined. In uranyl sulfate systems phase separations can occur in the temperature range of 250 to 300° C, although additions of sulfuric acid and alkali metal sulfates are effective in suppressing these. Dissolved copper is completely effective in recombining the hydrogen (or deuterium) and oxygen which are produced by the radiolytic decomposition of water in fissioning uranyl sulfate solutions. Alloys considered most as materials of construction for homogeneous reactors are Zircaloy-2, austenitic stainless steels, and commercially pure titanium. Austenitic stainless steels develop a protective anhydrous a Fe2O3 film at 250° C and higher temperatures. Stability of the film depends on solution composition, rate of flow and temperature. In the absence of an oxidant such as oxygen, uranyl ion is reduced and a very corrosive solution is formed. Crevices and stagnant regions where oxygen depletion can be encountered must be avoided. The out-of-pile corrosion behavior of stainless steel is relatively unchanged by exposure to a solution in which fission is taking place but where exposure is to solution circulating outside the neutron flux region. Zircaloy-2 is very resistant to attack in non-fissioning solutions. In-pile the corrosion rate increases with increasing power density and temperature. Titanium generally behaves like Zircaloy-2 but the in-pile corrosion rates are lower. Products accumulate in reactor solutions as a result of nuclear and corrosion processes and must be removed. The chemical behavior of some of the products has been investigated as have methods for their removal. Materials like iron, zirconium, barium, and the rare earths form insoluble salts or hydrolysis products which can be removed by filtration or centrifugation methods. Nickel, manganese, and cesium are soluble and are removed by solvent extraction or by separation of the uranium as a peroxide precipitate. Iodine, xenon, and krypton can be removed by stripping the fuel solution with oxygen or steam. Important progress has been made in developing thorium oxide slurry fuels for power breeder reactors. Thoria has been produced with controlled particle size, shape and surface area. The effects of firing temperature in the range of 450 to 1600° C, and the effects of pH and additives such as sulfate and silicate on the thoria properties have been investigated. Slurries have been irradiated at power densities and for times equivalent to those expected in a thorium breeder blanket without important change in properties. Molybdenum oxide is a promising catalyst for recombining radiolytic gas. Heat transfer and fluid flow characteristics in both laminar and turbulent flow

can be related to conventional correlations for Newtonian fluids through the non-Newtonian rheologic coefficients. Corrosion and erosion of structural materials are governed by the size and shape of the particles, the concentration and velocity and by the presence of additives. Data are being obtained to show the effects of radiation on corrosion. Thoria slurries have been circulated in engineering equipment for thousands of hours. Corrosion and erosion have been a problem only in the circulating pumps and in very high velocity or turbulence regions in the piping and valves. (A/CONF. 15/P/2391-Oak Ridge National Lab., Tenn.)

E-21. REACTOR TECHNOLOGY, PART II

Radioactivity Levels in Pressurized Water Reactor Systems. D. M. Wroughton and P. Cohen. 31 p. $0.50 (OTS). The three principal sources of radioactivity in the pressurized water reactor are discussed; activation of the water and its impurities by the neutron flux, activation of structures exposed to neutrons in the vicinity of the core, and fission products in the fuel. Information collected from plant operation and special tests is given on the radioactivity levels resulting from corrosion products and from fission product release through various modes of failure. Some observations on corrosion product and activity levels in operating systems are presented and, where possible, these levels are related to conditions of operation. (A/CONF. 15/P/410-Westinghouse Electric Corp. Bettis Plant, Pittsburgh.)

Damage Effects to Graphite Irradiated up to 1,000°C. R. E. Nightingale, J. M. Davidson, and W. A. Snyder. 13 p. $0.50 (OTS). Irradiation effects at 500° C. in experimental graphites with varied density, crystallinity, surface area, and pore distribution are discussed. Changes in macroscopic properties are dependent upon the initial crystallite structure; however, the mechanism by which the increase of interlayer spacing and decrease of apparent crystallite size effect these changes is not well understood. Macroscopic properties are also dependent upon the arrangement of crystallites and whole coke particles within the over-all structure which radiation may change slightly. Graphite irradiations in the Materials Testing Reactor extend exposure temperatures from 600 to 1,000° C. The temperature coefficient of property damage decreases with temperature, and only slightly less damage occurs at 1,000° C. then at 500° C. Thermal conductivity decreases by a factor of 50 with 30° C. irradiation, a factor of 3 with 500° C. irradiation, and a factor of 2 with 750° C. irradiation. Within the irradiation temperature range 500 to 1,000° C., the contraction rate after the first 1,500 Mwd/t is measurably the same: Co spacing expands slightly; and apparent crystallite size decreases by one-half. The total stored energy content is decreased with increased irradiation temperature. The way in which damage effects in irradiated graphite are distributed may be measured by the ease with which they may be thermally annealed. By estimating an activation energy for a given temperature-time annealing and measuring the amount of property annealed, at repeated increasing temperatures, it has been possible to characterize the damage with a damage distribution curve. Thermal annealing experiments on graphites with varied exposures and temperatures of irradiation have included property measurements of dimensions, thermal conductivity, interlayer spacing, and crystallite size. A damage mechanism is discussed which attempts to correlate and explain the changes in properties resulting from graphite irradiations at high temperature. Since the average crystallite does not contract (Co expands slightly) while the density of the graphite increases, reactor radiation at high

temperature must result in a more efficient packing of the crystallites. Damage distribution curves of samples irradiated at a variety of temperatures can be understood in terms of a process whereby energetic carbon atoms transfer a large amount of energy to the lattice close to a crystal defect. In this way, the activation energy for annealing which is in excess of equilibrium lattice temperature is supplied to the damaged lattice. (A/CONF. 15/P/614— General Electric Co. Hanford Atomic Products Operation, Richland, Wash.) Radiation Effects on the Oxidation Rate and on Other Chemical Properties of Graphite. G. R. Hennig, G. J. Dienes, and W. Kosiba. 12 p. $0.50 (OTS). Irradiation by fast particles changes most physical and chemical properties of graphite by producing various lattice defects such as displaced atoms, vacancies, dislocations, etc. Attempts have been made to evaluate the contributions of different types of defects to the property changes. Such evaluations have proven rather successful for physical properties and will be extended to several chemical properties. Experiments are described which show that oxidation catalysts and inhibitors act differently on irradiated and on unirradiated crystals. The oxidation of graphite crystals in ozone was studied because it appears to provide clues to the enhancement of oxidation rates in the presence of ionizing radiation. The kinetics and mode of attack of ozone on graphite crystals with or without prior neutron irradiation are described. (A/CONF. 15/P/1778-Argonne National Lab., Lemont, Ill.; Brookhaven National Lab., Upton, N. Y.; and General Atomics Div., General Dynamics Corp., San Diego, Calif.) Requirements and Techniques for Irradiation Testing of Reactor Materials for Pressurized Water Reactors. P. Cohen and D. M. Wroughton. 83 p. $0.50 (OTS). The problem of engineering tests of reactor materials and fuels are examined and illustrated in terms of the needs for and experience encountered in the development of the cores for pressurized-water reactors. The adaption of

facilities and test conditions to available test reactors is described along with the design problems of loops. The utilization of irradiation testing is illustrated by a brief discussion of the program and results for the development of uranium oxide fuel for use in the Shippingport Pressurized Water Reactor. (A/CONF. 15/P/2508-Westinghouse Electric Corp. Bettis Plant, Pittsburgh.)

E-22. REACTOR TECHNOLOGY, PART III

Thermal and Hydraulic Experiments for Pressurized Waters Reactors. W. H. Esselman, I. H. Mandil, S. J. Green, P. C. Ostergaard, and R. A. Fredrickson. 48 p. $0.50 (OTS). The results of an analytical and experimental program to obtain knowledge concerning thermal design unknowns for pressurized water reactors are described. Experimental techniques used are described. The out-of-pile tests simulated nuclear heating by using electrically heated test sections. The in-pile experiments were performed with specially instrumented and loaded fuel elements installed in a reactor to measure appropriate thermal and hydraulic conditions. Results on the following subjects are included: experiments on the surface effects of nucleate boiling; pressure drop tests; heat transfer burnout tests; transient loss-of-coolant-flow experiments with an operating reactor; and reactor performance during operation at limiting thermal and hydraulic conditions. (A/CONF. 15/P/453-Westinghouse Electric Corp. Atomic Power Div., Pittsburgh and Division of Reactor Development, AEC.)

Experimental Studies of Natural Circulation Boiling and Their Application to Boiling Reactor Performance. P. A. Lottes, J. F. Marchaterre, R. Viskanta, J. A. Thie, M. Petrick, R. J. Weatherhead, B. M. Hoglund, and W. S. Flinn. 33 p. $0.50 (OTS). A summary is given of the work carried out over the past few years in the field of experimental boiling fluid flow. The various apparatus tested are described along with pertinent variables investigated in each test. The tests described cover natural circulation boiling at pressures up to 600 psia, and power densities up to 95 kw/liter. Both steam-water densities and steam-water friction factors, for boiling with uniform heat addition, were studied. Also included are: (1) a brief description of boiling instability tests using electrically heated elements; and (2) a summary of visual studies using air-water mixtures at atmospheric pressure. Design equations, and hydrodynamic and heat transfer design criteria for natural circulation boiling reactors are presented. Included are methods for calculating steam volume fractions and recirculating flow rates as functions of reactor power. Heat transfer equations are listed, compared, and discussed. Burnout limitations are discussed. The method for calculating recirculating natural circulation boiling reactors is based on modifications of the standard equations for single-phase flow. (A/ CONF.15/P/1983-Argonne National Lab., Lemont, Ill. and Combustion Engineering, Inc. Reactor Development Div., Windsor, Conn.)

APPENDIX 14

CHIEFS OF UNITED STATES DELEGATION AND SECRETARIAT TO

SECOND INTERNATIONAL CONFERENCE

GENEVA, SWITZERLAND, SEPTEMBER 1-13, 1958

United States Representatives

LEWIS L. STRAUSS, United States Delegation Chairman; Special Assistant to the President on Atomic Affairs.

WILLARD F. LIBBY, United States Delegation Vice Chairman; Commissioner, Atomic Energy Commission.

JAMES R. KILLIAN, Jr., Special Assistant to the President for Science and Technology.

ROBERT MCKINNEY, United States Representative to the International Atomic Energy Agency, Vienna, Austria.

ISADOR I. RABI, United Nations Advisory Committee to the Secretary General on the Peaceful Uses of Atomic Energy; and Higgins Professor of Physics, Columbia University, N. Y.

Assistants to the United States Representatives

JOHN A. HALL, Director, Division of International Affairs, Atomic Energy Commission.

ALFRED IDDLES, President of the Atomic Industrial Forum, New York, N. Y. VIRGINIA WALKER, Office of the Special Assistant to the President on Atomic Affairs.

Special Advisers

Shields WarREN, United States Representative to the United Nations Scientific Committee on the Effects of Atomic Radiation; and Professor of Pathology, Harvard Medical School, Cambridge, Mass.

JOHN A. MCCONE, Chairman, Atomic Energy Commission.

Advisers and Members of the Congressional Joint Committee on Atomic Energy