The deposition of particles from turbulent gas streams due to thermal gradients was predicted using a simplified flow medel. In the medels particle concentration was assumed constant in the turbulent region. The particle deposition velocity was taken to be the radial velocity of thermal - repulsion in the laminar sublayer. Temperature gradients in the laminar sublayer were calculated for constant wall temperature using the Reynoldss analogy. The loss magnilude of suspended particles from loss due to thermal deposition is not generally greats and is most important for submicron particles. (auth)

The electrolytic oxidation in nitric acid of stainless steel, zirconium, Zircaloy-2, zirconium- uranium alloy, aluminum, and uranium - molybdenum alloy was demonstrated on a laboratory scale. The rate of chemical dissolution of UO/ sub 2/ in nitric acid was measured. Corrosion of stainless steel by these dissolver solutions was measured and found to be negligible. Electrolytic dissolution was demonstrated to be a practical technique for the first step in processing fuel elements of several types of power reactors. (auth)

In an attempt to determine the validity of the method of predicting the release of fission gases from U0/sub 2/ suggested recently by Cottrell et al., a series of calculations were made of the expected release of Kr/sup 85/ from prototype Experimental Gas-Cooled Reactor (EGCR) fuel capsule irradiated in the Oak Ridge Research Reactor (ORR). The computed values were then compared with measured values of the per cent Kr/sup 85/ released. In the calculations, the thermal conductivity of the U0/sub 2/ was assumed to be 0.028 w/cm- deg C in the temperature range from 700 to 1600 deg C, and in the absence of a precise knowledge of the helium gap, the cases of a 3-mil helium gas and no gap were treated. Values of the release-rate parameter (D) were estimated from BET surface areas of the U0/sub 2/ pellets. Results showed that the measured values of the per cent Kr/sup 85/ released generally fell within or close to the limits set by the 3-mil helium gap and no gap conditions. There was also a definite correlation between the measured values and the 3-mil gap condition when the clad temperature was about 700 deg C. When the clad temperature was …

An analog model of the EGCR steam generator was developed and operated on the ORNL analog computer as part of a program to simulate the operation and control of the EGCR reactor plant. Equilibrium operation and the transient response of the steam generator unit to system perturbations were studied. A simultaneous solution of the basic heat transfer equations representing the performance of the unit was obtained. The model was operated initially at steady- state conditions, and then perturbations were made to gas flow, gas inlet temperature, and steam throttle valve position. The response characteristics of the model during the transients were recorded. The steam generator gas outlet temperature showed a marked degree of insensitivity to changes in gas inlet temperature. The effect of gas flow changes on gas exit temperature was slightly more pronounced. The transient behavio-r of the unit was reasonable, and the model developed indicated satisfactory operation within the design range of 20 to l00% of full power. (auth)

An economic analysis is presented for the fabrication of replacement cores for SM and PM type reactors, including analysis of various core types and core fabrication technologies. The analysis indicates that major savings are possible by utilizing Type 3 cores (40-mil plates, 25 wt% UO/sub 2/, welded assembly) in all SM and PM type reactors, and that significant savings are possible by multiple core procurement and reprocessing, and relaxation of cobalt and tantalum requirements in Type 347 stainless steel. (auth)

References (297) were obtained from Applied Science and Technology Index, Chemical Abstracts, Dissertations in Physics, Nuclear Science Abstracts, and Science Abstracts, Section A. The period covered was from 1948 through l960. The arrangement is alphabetical by title; personal author and subject indexes are included. (P.C.H.)

Research concerned with a theoretical and experimental study of magneto- thermionic power generation is reported. This concept promises a number of advantsges over conventional thermionic generators. Low pressure Cs diodes suffer from an undesirable size limitation, because selfinduced magnetic fields reduce current transmission and hence, efficiency. Although this effect does not arise in the small devices tested to date, it becomes important in building larger generators. Analysis indicates that this problem can be virtually eliminnted by the application of a longitudinal magnetic field. An experimental and theoretical study was initiated to verify the predicted magnetic field effects, and to examine their possible use in a-c generation. (auth)

Methods were developed for analysis of cooling water for impurities, radioisotopes, etc., and experimental results are presented for the ORNL Research Reactor. The theory of nuclear reactions in a water-cooled reactor is discussed at length, and equations were developed which allows predictions of equilibrium conditions from nonequilibrium measurements. The equations were verified experimentally by work on the ORNL Research Reactor and can be extended to other reactors. The origins of Na/sup 24/, Cd, and fission product activities are discussed, and the possibility of fuel element rupture detection by delayed neutron measurements is considered. (D.L.C.)

Various reactor concepts were compared for use as a separate superheater which could be added on to an advanced 300-Mwe reactor producing saturated steam. Fossil steam plant superheat temperatures were used as a criterion for selecting nuclear superheat temperatures. Therefore, the performance specified for the superheater was a minimum exit steam temperature of 566 deg C (1050 deg F) when supplied with saturated steam at either 7l atm (l050 psia) or l67 atm (2450 psia). A preliminary screening of ten different reactor concepts resulted in the selection of two for a detailed evaluation. These are a direct-cycle, watermoderated reactor, and an indirect-cycle, sodium-cooled reactor. The steam- cooled, water-moderated system is judged to have the best chance for initially reaching 566 deg C (l050 deg F), whereas, the indirect-cycle, sodiumcooled system is considered best for subsequent advances to exploit the more efficient, high- pressure steam reheat cycles. A design concept was selected for each of the reactors to establish a basis for the detailed evaluations and comparisons. The technical evaluation of the two concepts shows that the sodium-cooled, indirect- cycle superheater can realize a significant reductlon in the gross plant heat rate STA0.6097 to 0.5404cal/watt-sec (87l0 to 7720 Btu/kwh)! by use …

None

Progress is reported in the development and evaluation of reactor materials and components, fuel and fuel-element development studies, growth of UO/ sub 2/ single crystals, radioisotope and radiation applications, corrosion studies of the fluoride-volatility process, assistance to the CANDU Reactor program, radiation effects studies of MGCR and SM-2 fuel materials, gas-cooled reactor program developments, corrosion of thorium and uranium under storage conditions, and gas-pressure bonding of beryllium-clad elements. (B.O.G.)

Changes in the mechanical properties of plain carbon and alloy steels that are caused by fast neutron irradiation are presented in graphic form. These data were abstracted from classified and unclassified reports published since 1948 by USAEC, AECL, and AERE. Data are included for the following steels: A- 105-55T, A-106, A-200-55T-T22 (Croloy), A-201, A-242, A-301-B, A-302-B, AlSI-C- 1013, AISI-C-1065, SA-70, SA-193,4, SA-212, SA-285, SA-336, SAE-1018, SAE-1020, SAE-1042, SAE-1045, SAE-1095, SAE-1113, SAE-1141, SAE-4130, and SAE-4340. The mechanical properties for which data are reported include hardness, yield strength, tensile strength, total and uniform elongation, reduction of area, fatigue strength, notch factor, creep, impact energy, and transition temperature. (auth)

The effect of temperature on the unnotched and notched Charpy impact properties of plutonium was studied in the alpha, beta, gamma, and delta phases encompassing a temperature range of -43 to 330 deg C. Impact energies for unnotched specimens generally increased with increasing test temperature in the alpha and beta phases. Brittle failures were obtained in these phases. The specimen tested in the gamma phase did not fracture but bcnt in ofi U-shape and pulled through the anvil. lmpact energies for the Charpy V-notched specimens were much lower than for the unnotched spcc(mens.dt corresponding temperatures. Brittle failures were obtained in the alpha and beta phases; some ductility was evidenced in the gamma phase while the specimen tested in the delta phase did not fracture but bent in a U-shape. It was concluded that the beta phase is quite sensitive to both strain rate and notch effects. Fracture appearance in the alpha and beta phase is discussed from the standpoint of grain boundary effects and microcracking. (auth)

A manipulator tong was modified to include an alpha seal . at the slave end. This arrangement is used in conjunction with manipulators in a lead- shielded glove box to obtain protection from both gamma and alpha radiation (auth)

Developments in chemical engineering are described. Pyrometallurgical development, fuel-processing facilities for EBR-II, pyrometallurgical research, laboratory and engineering-scale investigations of fluoride volatility processes, conversion of UF/sup 6/ to UO/sup 2/, calcination studies in small-diamenter columns, metal oxidation and ignition kinetics, metal-water reactions, determination of nuclear constants, thermally regenerative emf cell, thermoelectricity research, reactor decontamination, waste processing, and the high-level gamma-irradiation facility are discussed. (M.C.G.)

Tests were carried out to determine periodically the radiation level at all monitored points in the radioactive waste disposal plant and the effictiveness of shield facilities in working areas at times during which the radiation level is above limits set by Health Physics. The radiation surveys were made using radiation control instruments available on the site. Air samplings were made, as necessary, to reflect activity level changes which occurred during incinerator or drumming operations. The radiation control measures in effect during the period of test performance appeared to be adequate. No conclusion could be made as to the effectiveness of the shielding facilities in the work areas. (M.C.G.)

Tests were conducted to determine the behavior of lithium hydride and calcium hydride under a variety of conditions. Metals of interest as container materials were subjected to stress-rupture and tensile tests after undergoing one or more thermal cycles to certain maximum temperatures while in contact with lithium hydride at various lengths of time. Calcium hydride was produced and fabricated, then tested to determine some of its physical properties. (auth)

A neutron absorption cell was built to monttor continuously the boric actd concentration of reactor water flowing through a purtfication loop. The instrument was calibrated over the range from 0 to 2.65 g/l with a maxtmum absolute error of 0.05 g/-l. (auth)

Elementary-particle data and other reference information for use in high- energy physics are presented in tables and graphs. (M.C.G.)

Coolant flow for gas-cooled in-pile loops must be supplied during irradiation test runs. A centrifugal compressor has been designed and developed for circulating helium at volume flows from 75 to 250 acfm at compressor suction conditions of 400 psi and 600 deg F. The compressor using grease-lubricated ball bearings has operated satnksfactorily for a total of 3500 hr. (auth)

Integral diffusion coefficients, viscosities, densities, and surface tensions were measured for the aqueous uranyl nitrate --nitric acid--aluminum nitrate system. The variation of the uranium diffusion coefficient was determined as as a function of uranium concentration, nitric acid concentration, and temperature. (auth)

Macro- and high-resolution methods for autoradiography were used to investigate a number of problems concerning the distribution of uranium and other active elements in metal specimens. Illustrations are presented of the behavior of some of these materials with various henting and cooling treatments and in the presence of metals with which intermetallic phases are formed. Techniques are discussed for handling and examining autoradiographic stripping films by means of various types of metallurgical and conventional microscopic equipment. (auth)

An evaluation was made of the hazards associated with studies of the use of U/sup 235/ as fuel in the Fast Reactor Zero Power Experiment (ZPR-III) in the design of EBR-II. The systems are blanketed cores having a range of compositions in which the volume fractions of U/sup 235/, U/sup 238/ Fe, and Na are varied. Methods used for protection against accident or damage are described. Excursions resulting if the safety mechanisms did not operate are discussed. (M.C.G.)

The heat capacity of terbium metal was measured between 0.25 and 1.0 deg K. The high temperature data can be expressed as C = 28R x 10/sup -3/T/sup 2/ cal/ mole-deg. The data are compared with a Schottky formula for the heat capacity, assuming equal spacing between the nuclear hyperfine levels with an over-all spacing of beta = 0.45/ deg K, correspondlng to the high temperature expression above. The agreement is within experimental error except at the very lowest temperatures obtained. The effective field at the nucleus of the terbium atom can be calculated and is found to be H/sub eff/= 4 x 10/sup 6/ gauss. (auth)