I. An adequate and conservative calculational method for evaluation of the heat generation distribution in the primary sodium system substructural areas was developed. The method was programed for the IBM 704 and the IBM 709. The results obtained from analysis of the gamma heat generation in the primary sodium pipe tunnels and in the intermediate heat exchanger cells are presented. Calculations are outlined, and gamma attenuation coefficients for concrete, sodium, and steel are given. II. Results obtained from analysis of the gamma heat generation in areas where the primary sodium system piping layout was changed from that of the previous analysis are presented. Major changes in magnitude of the hot spot heat generation due to the changes are pointed out. (auth)

The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)

The following criteria outlines the basis, objectives, and fundamental methods that shall govern the preparation of final design for ``Project CGI-839, Modification to Fuel Element Test Facilities -- 1706 KER.`` These modifications will provide the equipment to test NPR size fuel elements in the KER recirculating loops. The 1706-KER Recirculation Test Facility of KE Reactor is operated to obtain experimental data regarding high temperature reactor coolant technology and high temperature fuel element testing. The facility consists of four in-pile recirculating loops. These loops will permit testing of fuel elements with the existing process tubes of 2.1 inches I.D. To provide adequate in-reactor fuel element test facilities to support the development of NPR fuel, two KER loops, {number_sign}3 and {number_sign}4 will be converted to provide a process tube of 2.7 inches ID that will be operated at typical NPR irradiation conditions. The remaining loops No. 1 and 2, will be modified to provide additional flow and heat transfer capacity for greater flexibility in the testing of high temperature fuel elements smaller than the NPR size. New pumps, heat exchangers, and minor piping modifications will be required in all loops.

BS>A previous calculation of the rate of diffusion of like charged particles across a magnetic field is generalized. No "a priori" assumption as to the relative magnitude of certain terms need be made and spatial density gradients are permitted in both directions perpendicular to the field. The final result agrees with that given earlier. (auth)

Economic aspects of nuclear power development in the U. S. are tabulated and graphed. Included are figures on presently operating reactors as well as those contemplated or scheduled. Also a brief description of the objectives of short- and long-range programs is given as well as tables listing some of the characteristics of each reactor. (J.R.D.)

A study was made of the effect of carbon, manganese, oxygen, palladium, and sulfur on the structure and fabricability of iron-35 wt.% chronium-1 wt.% yttrium alloy. Using a vacuum-induction melting technique each of the additives except oxygen was introduced to 1-lb remelts of a single 15-lb master alloy. The master alloy and remelts were made under similar melting, pouring, and casting conditions. Oxygen was introduced as Fe/sub 2/0/sub 2/ by inertelectrode arc melting to avoid extraneous, uncontrolled contamination stemming from crucible contact. Photomicrographs were prepared of as-cast metal illustrating structural variations. Ingots obtained were fabricated to 0.050-in. sheet at 2000 deg F to compare fabrication characteristics with those of the control ingot containing no additives. As a qualitative measure of metal soundness and ductility, a portion of each of the 0.050-in. sheets was further reduced at room temperatare to 3-mil foil. During melting at 2900 to 3000 deg F under controlled conditions, the amount of yttrium present in the charge was reduced roughly 50% by reaction with the alumina crucible. Sulfide, added as FeS, and oxide additions also lowered the amount of yttrium retained in ingots. The maximum amount of sulfur retained in an alloy of nominal composition iron-35 wt.% …

This report discusses a proposal to measure the corrosion rate of aluminum-clad DOE elements in deionized water in one of the KER Loops at the highest obtainable surface temperatures under conditions where the bulk water adjacent to the elements is saturated with aluminum corrosion product. This will determine whether in-reactor corrosion may be reduced by controlling the aluminum concentration in the water and also provide corrosion data in deionized water at higher temperatures than is now available. A thermocouple slug will be included.

The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum- 35 wt.% uranium were investigated. Initial investigations were concerned with the effects of 3 at.% ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 wt% level, additions of germanium, silicon, tin, or zirconium inhibited the formation of UAl/sub 4/ and thereby increased the extent of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl/sub 4/ and free lead was detected in the alloy to which lead had been added as the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of rior when evaluated on the bases of casting qualities and recycling characteristics. (auth)

Abstract: "The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum-35 w/o uranium were investigated. Initial investigations were concerned with the effects of 3 w/o ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 w/o level, additions of germanium, silicon, tin or zirconium inhibited the formation of UAl4 and thereby increased the content of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl4 and free lead was detected in the alloy to which lead had been added at the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of aluminum-35 w/o uranium alloys, and may prove superior when evaluated on the bases of casting qualities and recycling characteristics."

Zircaloy-2 is anodically converted to scaly ZrO/sub 2/ at 60 deg C in 8 M HNO/sub 3/. About 0.5 mole of acid is consumed per faraday, and after saturation of the electrolyte with nitrogen oxides about 0.3 mole of gas is evolved per faraday. The nitric acid is reduced to hydrogen, NO, and N0/sub 2/, with hydrogen predominating if the cathode is Zircaloy and NO if the cathode is platinum. Corrosion specimens of HRT metals were exposed to the electrolysis conditions. From determinations of the decomposition potential of nitric acid it appears that a metal container for the electrolytic process can be protected from stray-current corrosion by holdlng it at a potential --0.5 volt positive to a platinum cathode operating at a current density of 5 to 10 ma/cm/sup 2/. Practical laboratory experiments tended to confirm this conclusion. (auth)

Pressed sulfur pellets, irradiated for two hours at 10 w yielded satisfactory counting rates on an internal proportional counter for measuring fast neutrons. The pellets were counted in their original form and an equation applicable to thick sources was applied to obtain disintegration rates. Decay data and beta energy analyses indicated that the induced activity was due to P/ sup 32/. Cross section values reported in the literature for the S/sup 32/(n,p)P/ sup 32/ reaction involved were found to differ widely. The fust flux values given are based on an assumed cross section of 285 millibarns in a fission spectrum. Thermal flux values at the same positions and for the same power are also presented. (auth)

The flooding characteristics of fluorothene Raschig rings at low liquid rates with the existing literature data on carbon rings are compared. Visual observations of flooding in the air-water system were correlated with pressure drop using a 3-in. i.d. glass column having an 18-in. long fluorothene Raschig ring packed section. A comparison of extrapolated literature data with these experimental data indicates thc gas flow rates required to flood fluorothene packing are slightly higher than for carbon packing. At a liquid flow rate of 537 lbs/hr-ft/sup 2/, fluorothene rings flooded at a gas velocity of 842 lbs/hr- ft/sup 2/ whereas carbon rings are reported to fiood at a gas velocity of approximately 755

A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)

The refractory uranium compounds UC and UC/sub 2/ were prepared by arc- melting methods, UN by gas solid reactions, and UB/sub 2/ and UBe/aub 13/ by solid-solid reactions. Techniques for sintering powder compacts of these compounds to densities 90% of theoretical or greater were developed. Mean linear thermal-expansion coefficients of 7.9 x 10/sup -6/ per deg F for Uc and 9.3 x 1O/ sup -6/ per deg F for UBe/sub 13/ were obtained in the range 68 to 1800 deg F. The thermal conductivity of sintered UBe/sub 13/ increased from 0.050 cal/ (sec)(cm)(C) at 1OO deg c to 0.064 cal/(sec)(cm)(C) at 650 deg C. The electrical resistivity of sintered UBe/sub 13/ was 113 microhm-cm at 27.6 deg c. Knoop microhardness values varied from 500 for sintered UC/sub 2/ to 1155 for sintered UBe/sub 13/. Sintered Uc and Uc/sub 2/ decompose rapidly in boiling (atmospheric pressure) water. A sintered UBe/sub 13/ compact lost 2.2 mg per cm/sup 2/ (average), after 72 hr in boiling water (atmospheric pressure). Uc/sub 2/ reacts more rapidly with nitrogen and oxygen than does pure uranium. UB/sub 2/ and UBe/ sub 13/ were less reactive with oxygen and nitrogen than is pure uranium. The heat of formation …

The temperature characteristics associated with the handling of self-boiling radioactive wastes from the separations extraction processes in the Chemical Processing Department have necessitated several heat transfer studies. Earlier studies 1,2,3 defined the feasibility of self-concentration in existing waste storage facilities by determining the rate of heat generation from the decay of stored fission products and by defining the rate of heat loss from existing storage tanks to the surrounding soil.

S>The results of all laboratory research and development on the process for separation of barium-140 from MTR fuel elements are presented. The steps include caustic dissolution separation of barium and strontium with fuming nitric acid and removal of strontium by the chromate-acetate method. The results of laboratory and pilot plant corrosion investigations and high radiation level flowsheet tests in the Multicurie Cell are also included. ( auth)

Tests were conducted to determine the limit of error of the temperature measuring system for high thermal-stress tests on moderator cans and to determine a means for the calibration of chromel-alumel thermocouples after installation. Results and recommendations are included. (J.R.D.)

The following report discusses the construction of two heavy-ion linear accelerators, one at the the University of California Radiation Laboratory and the other at Yale University. The goal for these accelerators is to produce ions of masses up to that of neon with energies of 10 Mev per nucleon.

Introduction: This document describes the results of a six-week conceptual design study of an organic moderated and cooled reactor (OMCR) power plant adapted to a Class T-7 tanker.

A micro-electrephoresis cell is described, and its application to the determination of the zeta potential of thorium oxide is presented. Samples of thorium oxide from different sources, some of which were subjected to certain physical treatments, are charcterized by the zeta potential obtained in water. The zeta potentials produced in solutions of HCl, H/sub 2/SO/sub 4/, KOH, NaOH, Na/sub 4/P/sub 2/O/sub 7/ and Na /sub 3/PO/sub 4/ are also given

Critical assembly studies were conducted tc provide physics and engineering data to aid in developing the Mobile Low-Power Reactor (ML-1). The ML-1-lA core was critical with 59 elements containing 17,906.71 g of U/sup 235/ and had an excess reactivity of 0.381 x 10/sup -2/ DELTA k/k at a moderator temperature of 24.91 deg C. The ratio of maximum element power to core-averaged power was approximately 1.09. The ratio of maximum to core-averaged thermal flux was approximately 1.10. At an 18-deg separation, the shutdown worth of the cadmium-covered control-blade mock-up was 1.14 x 10/sup -2/ DELTA k/k for a 69 element core. Radial and upper axial reflector-moderator void coefficients were - 0.59 plus or minus 0.07 and -0.36S plus or minus 0.015 x 10/sup -2/ DELTA k/ k.per in., respectively. Two lA production fuel elements were evaluated in the critical assembly core. The results predict that the production elements tested contained roughly the same fuel as the critical assembly element and an additional 772 g of stainless steel equivalent on the average. Radial power and neutron flux distributions were measured in a 19-pin lB fuel element. Fairly uniform distributions were observed. Data to evaluate the thermal utilization of this element were …

A series of neutron multiplication measurements with arrays of 1 by 8 by 10 in. slabs of 93.4% U/sup 235/-enriched uranium metal was made to provide data from which safety criteria for the storage of these fissile units can be established. Each slab contained 22.9 kg of U/sup 235/. A maximum of 125 units was assembled. The arrays studied were cubic lattices of the units and were usually parallelepipedal in shape. Arrays were both unmoderated and Plexiglas- moderated and were surrounded in most cases by a 1-in.-thick Plexiglas reflector. The lattice densities (ratio of fissile unit volume to lattice cell volume) were between 0.023 and 0.06. Unmoderated lattices with a density of 0.06 would require 145 plus or minus 5 units for criticality, while those with a density of 0.023 would require 350 plus or minus 30 units. In lattices in which the fissile units are separated by 1 in. of Plexiglas, approximately 27 units would be required for a critical array with a lattice density of 0.06 and about 75 units for a density of 0.023. Distributing Foamglas (containing 2% boron) throughout a moderated array increased the critical number of fissile units by a factor of 5, while Styrofoam …

A series of neutron multiplication measurements with arrays of 1 by 8 by 10 in. slabs of 93.4% U235-enirched uranium metal have been made to provide data from which safety criteria for the storage of these fissile units can be established. Each slab contained 22.9 kg of U235. A maximum of 125 unites was assembled. The arrays studied were cubic lattices of the units and were usually parallelepipedal in shape.

Abstract: "The niobium-oxygen equilibrium has been determined by metallographic examination of arc-cast alloys made of electron-gun-refined niobium metal and special purity niobium pentoxide. Two intermediate oxides, NbO and NbO2, melt without decomposition at 1945 C and 1915 C, respectively. Eutectic reactions exist between niobium and NbO and 1915 C and between NbO and NbO2 at 1810 C. Experimental supports a peritectic reaction between NbO2 and Nb2O5 1510 C. The maximum solid solubility of oxygen in niobium metal is 0.72 weight per cent."