Activity and ionic impurity specifications are presented for Thorex thorium nitrate products. Two sets of specifications are given, one set for direct handling during refabrication of production reactor thorium metal slugs and the second for refabrication of future power reactor thorium metal elements by semi-remote technics. Consideration was given to the health hazard problems associated with each process step between the Thorex process and final refabricated source material in order to arrive at these specifications.

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on circulating-fuel reactors and ether ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided into three major parts: I. Reactor Theory, Component Development, and Construction, II. Materials Research, and III. Shielding Research.

A new gas-liquid countercurrent system (the ANCO system from Anisole-Complex) for the enrichment of boron isotopes has been developed. It is believed that use of this systems will result in a considerably lower unit cost for enriched boron-10 than was previously possible. The system utilizes the exchange reaction between BF3 (gas) and BF3·anisole (liquid) to concentrate boron-10 in the liquid phase. The single stage isotopic separation factor for this system has been found to vary from 1.039 at 0°C to 1.029 at 30°C. The isotopic exchange reaction has been shown to be rapid. Vapor pressures of the complex as a function of temperature have been measured and the heat of formation of the complex determined. Laboratory experiments show that quantitative removal of the BF3 from the complex can be accomplished by heating. A complete miniature ANCO plant was constructed and operated in the laboratory to test the feasibility of the system. The system was found to operate efficiently with a minimum of attention, and to enrich the isotopes of boron as expected. Based upon the experience obtained with the laboratory ANCO unit, a pilot plant large enough to utilize a 6-inch diameter exchange column was designed. The design calculations of …

The ORNL Graphite Reactor operated very well during 1955. The downtime was low, only 8.6%. The fuel in the bonded slugs did not perform as well in 1955 as in 1954. Much of the trouble was undoubtedly due to growth of slugs which were not beta-transformed. It is known that some slugs had grown over 1/2 in. The automatic central system installed in 1954 continued to operate satisfactorily. The cooling system gave minor trouble when one of the 900-hp fan meters had to be replaced because of shorts in the rotor. The high radiation in the canal was the largest source of trouble. Approximately 55 tons of slugs discharged from the reactor in 1952 was sent to the Metal Recovery Plant. Enough slugs had raptured, due to their long exposure in the canal and reactor, to badly contaminate of water. Most of the contamination was removed by the end of the year, but the radioactivity which had soaked into the canal wells was enough to give high radiation fields. A solution to this problem was being sought at the end of the year. A study is under way on the possibility of increasing the flux of the ORNL Graphite Reactor …

Progress report of the Oak Ridge National Laboratory Metallurgy Division providing updates on various projects, experiments, and other work. This report includes a summary of scholarly output and summaries of general corrosion, dynamic corrosion, mechanical properties, nondestructive testing, welding and brazing, fabrication, physical chemistry of corrosion, HRP metallurgy, the package power reactor program, APPR absorber, metallurgical materials and processing, metallography, ceramics, fundamental physico-metallurgical research.

The Thorex Pilot Plant at Oak Ridge National Laboratory was operated during 1955, processing reactor-irradiated thorium slugs to recover U233 and thorium and 12 MTR fuel elements to recover U235 and Np237. The radiation exposure received by operating personnel during this period averaged 60 mrcp/man-week. Most radiation exposure was received in areas that were intended to be only slightly or nonradioactive. However, because insufficient decontamination of process solutions was achieved and equipment surfaces became contaminated from equipment failures, these areas became primary sources of personnel exposure. The installation of additional shielding where needed and the prompt removal of surface contamination successfully reduced the radiation levels and exposures in these areas. Remote control of processing equipment and sampling of very radioactive solutions from process equipment was successfully accomplished, and assisted in the reduction of exposure to operating personnel.

Progress report of the Oak Ridge National Laboratory Aircraft Nuclear Propulsion Project providing updates on various projects, experiments, and other work. This report includes summaries of project activities in: aircraft reactor test design, ART physics, ART instruments and controls, component development and testing, procurement and construction, ART, ETU, and in-pile loop operations, phase equilibrium studies, chemical reactions in molten salts, physical properties of molten materials, production of fuels, compatibility of materials at high temperatures, chemistry, analytical chemistry, metallurgy, dynamic corrosion studies, general corrosion studies, fabrication research, welding and brazing investigations, mechanical properties studies, ceramic research, nondestructive testing studies, heat transfer and physical properties, radiation damage, fuel recovery and reprocessing, critical experiments.

A partial bibliography of reports on the Fluorox moving-bed for producing UO3, UF4, and UF6 is given.

Calculated data and graphs describing the effects of batch thermal-neutron irradiation of thorium, the usual method of operation of heterogeneous reactors, are presented. The buildup and decay of U233, Pa233, other heavy isotopes, and fission products are considered on the basis of best available cross-section and fission-yield data. The effects of these irradiation products on the Thorex chemical separation process are indicated briefly.

The Homogeneous Reactor Test (HRT) is the experimental reactor facility (Frontispiece) being designed and constructed at ORNL as the next step in homogeneous reactor development between the 1-Mv HRE and a "full-scale" power station. The HRT will provide an integrated test at 5 to 10 Mv for the flowsheet and equipment designs on which the full-scale effort will be based. Furthermore, its design is such that several homogeneous systems which require essentially the same operating equipment may be tested with comparatively minor modifications of the original reactor installation. The reactor will be assembled in the building which housed the HRE, located in the experimental reactor exclusion area approximately one mile south of the oak ridge laboratory. (See figure 1) / It is the purpose of this report to provide information with which the hazardous aspects of this reactor may be evaluated. Briefly, it will be shown after a statement of purpose and a general description of the reactor that: 1. The design characteristics and equipment requirements are such that escape of highly reactive material from the reactor piping is unlikely. 2. Should the entire core and blanket contents suddenly escape from the reactor system, a seal-welded steel tank surrounding the …

The flux through a given region is proportional to the total lengths of the neutron flight paths that intersect that region. The analytical Monte Carlo procedure manufactured neutron flight paths and totaled the lengths of all paths intercepted by the regions illustrated in Figure 1. The procedure was designed to utilize the various symmetries in the lattice. / Consider a portion of the lattice whose planar cross-section is shown in Figure 5. If R is the region in which the flux is to be estimated and F the fuel rod in which the neutron originated, then flight path P results in an intercepted length whose reflection in the plane is L. On the other hand flight path P' intercepts R' with length L'. R' is not the region to be studied, but a translation of the flight path P' to F' would result in the neutron intercepting R. The origin in P was arbitrary. For each neutron originating in P another could, with equal probability, have originated in P' with parallel paths. Hence consulting L' in R' towards the total flux is equivalent to starting a neutron at P'. Thus consideration of all regions symmetric to R with respect to …

This semiannual progress report and future reports will be published as two documents to permit a wider distribution of the unclassified material. The report numbers are assigned in sequence so that the two reports will fall together when filed by report number.

Methods for the determination of plutonium and uranium in highly radioactive scrup dissolver solutions have been developed. Plutonium was separated from the dissolver solutions by solvent-extraction and ion-exchange techniques and determined by potentiometric titration. Uranium was separated by ion exchange and determined by potentiometric titration. Solutions that were similar to the actual dissolver solutions and that contained known amounts of plutonium and uranium were analyzed by these methods. Evaluation of the data secured for the determination of plutonium and uranium by the methods given herein indicated that, within the limits of the precision of the methods, there was no bias. The precision of the data obtained for the determination of plutonium, expressed as the relative standard deviation, was better than 2% for plutonium in the concentration range of 0.27 to 0.64 mg/ml. The precision for uranium was estimated to be about 0.2% for uranium concentrations of 425 mg/ml. These methods and the data obtained by then are discussed in this report; the procedures are appended.

The Aircraft Reactor Experiment (ARE)was successfully concluded in November of 1954, and a detailed report of the operation was published the following year. At that time it was thought that an extensive examination of the reactor and system components after disassembly was warranted. It was realized, of course, that the level of radioactivity of the components would necessitate extensive delays in the examinations. Since examination of a few critical ARE samples showed nothing unexpected, much of the planned hot-cell inspection was postponed and complete examination of all but a few specimens was indefinitely suspended. The few examinations that were completed are described in this report, along with a description of the disassembly of the ARE system. Diagrams of the fuel system, sodium system, and off-gas system are presented in Figs. 1, 2, and 3 for reference use in visualizing the disassembly process.

New systems involving the exchange of boron between boron trifluoride and boron trifluoride addition compounds have been explored. These systems have large separation factors and potentially simple reflux mechanisms. A precise determination of this separation factor for the anisole-boron trifluoride system gave the value (see report). Boron exchange was found to occur between BF and BCl3. Several homogenous catalysts have been found which activate the hydrogen-water exchange, but none are adoptable to the production of deuterium because of the slow exchange rate. Platinum or platinum oxide may be usable as a heterogeneous catalyst with proper support or dispersion techniques. The high-pressure solubility of hydrogen in several amalgams was investigated in connection with a unique countercurrent exchange system. A proposed system involving isotopic exchange between lithium dipivaloylmethane in diethyl ether and lithium hydroxide in aqueous solution was shown to give little or no isotopic separation. Column studies of the carbonate system exchange reaction were concluded with a 40°C run. Slightly higher enrichment of N15 was obtained than at 30°C . The temperature dependence of all in this system was measured between 15 and 45°C. The factor increases with temperature, showing a tendency toward a maximum near 45°C. Isotopic exchange appears to …

The development of ionic methods for the determination of corrosion products in the highly radioactive Homogeneous Reactor (HR) fuels has been of major interest in the work of the Ionic Analyses Laboratory. Methods for the spectrophotometric determination of aluminum and for the polarographic determination of iron in HR fuels have been developed. The polarographic determination of molybdenum in uranyl sulfate solutions was studied. A polarographic method for the determination of zinc was developed. A fluorometric method for the determination of microgram amounts of fluoride was studied. Three organic reagents were investigated as precipitants for microgram quantities of zirconium in HR fuel. The automatic photometric titration technique was applied to the determination of thorium and of sulfate. A method was developed for the ionexchange separation and potentiometric titration of cobalt. The ultraviolet absorption spectra of technetium and rhenium were studied.

A preliminary chemical flowsheet is presented of a fluoride volatility process for recovering and decontaminating uranium from heterogeneous reactor fuels after dissolution in a fused salt. In laboratory work, a gross β decontamination factor of > 10 4 was obtained in the fluorination of a UF4-NaF-ZrF4 melt by passing the product UF6 through NaF at 650°C. The solubility of UF6 in molten NaF-ZrF4 was shown in kinetic studies to cause a lag in the evolution of UF6 from the fluorinator. Corrosion of nickel in the fluorination step appeared to be 2-4 mils/hr during the time that uranium was present. The average corrosion rate over the process as a whole was less than O.4 mil/hr. Earlier studies were reported in ORNL-1709 and 1877.

A theoretical and experimental investigation of the coefficient for diffusion of ions across a magnetic field Is described. The resultant diffusion coefficient is found to vary inversely as the square of the magnetic field strength, in accord with the usual collison-diffusion theory. The magnitude of the coefficient is much larger (x700) than the coefficient predicted by the usual ambipolar diffusion theory. This discrepancy is resolved by showing that diffusion across a magnetic field is not ambipolar in character in most arc experiments. The final experimental and theoretical values are in good agreement, and it is unecessary to postulate any additional diffusion mechanisms, such as plasma oscillations.

The development of the reactor layout is continuing. New features that have been incorporated because of stress, fluid flow, or fabricability considerations include an elliptical fuel expansion tank, a rounded dome to enclose the top of the reactor, a newly designed sodium pump impeller, and other related items. Recently completed heat exchanger tests yielded consistent data from which a series of heat exchangers is being designed. The most promising of these will be chosen for the ART.

A summary of the laboratory development program on aqueous uranium slurry fuels for the Homogenous Reactor Project during the period April 1951 through March 1953 is presented. These investigations were devoted primarily to a study of the uranium oxides in aqueous suspensions. It was concluded that U(VI) was most likely to be the stable valence state in such slurry fuels and it was shown that β-UO3·H2O platelet crystals were the stable modification at 250°C. Very pure slurries of β-UO3·H2O platelets, uranium concentration of 250g/liter and average particle size of about 10 μ, had favorable settling rates and could be easily redispersed. Their viscosity and corrosion rate in stainless steel were comparable with those in water. Exposure of these slurries to pile radiation disclosed that radiolytic hydrogen and oxygen gas pressure comparable in magnitude to those of uncatalyzed uranyl sulfate solutions could be expected. Fission products in the irradiated slurries were predominantly associated with the solids. Radiation also tended to promote caking of these solids on the walls of the radiation bombs. Uranyl phosphate and the magnesium uranates were briefly investigated as alternate system but were not found satisfactory. The program was discontinued before the feasibility of uranium slurries for reactor …

An investigation of the materials for use in the water-moderated and cooled Aray Package Power Reactor (APPR) operating at about 500°F was made. The available literature was analyzed, and the results of the different investigators were compared and averaged. Twenty different materials, including stainless steels, nickel alloys, Stellites and others, were investigated from the point of view of physical properties, susceptibility to radiation damage, and corrosion resistance. Corrosion rates were established for all the materials under various conditions, such as irradiation, flow, weld, stress, and various water conditions. Type-304 stainless steel was selected as the basic structural material. Operating conditions, to maintain minimum corrosion, were established also.

The fabrication of heat exchangers and radiators to be used in conjunction with high-temperature nuclear reactors may present exceedingly complex problems. Rigid heat transfer requirements may necessitate the use of compact assemblies of thin-walled small-diameter tubes as integral parts of the heat transfer units. Intricate designs may also be required in which cooling fins must be securely joined to the tubes at closely spaced intervals. In addition to the difficulties in fabrication imposed by the designs themselves, the high operating temperatures involved require the careful selection of materials and joining techniques. The choice of fabrication procedure for a given component must not only be based upon the stresses and temperatures to be encountered, but also upon special factors peculiar to nuclear service. Since many reactor applications employ highly corrosive environments, compatibility of the structural ma terials with the corrosive media is of paramount importance. The low nuclear cross-section require ment for brazing alloys to be used inside the re actor also places stringent limitations on the possible choices of in-pile applications. The use of boron in alloys for certain service may not be considered feasible, for example, because of its high nuclear absorption cross section. Although welding is used extensively …

Continued work on the adsorbability of metal complexes from concentrated LiCl solutions and LiCl-HCl mixtures on a strong-base anion-exchange resin further demonstrated the much higher adsorbability of these complexes from LiCl solutions than from HCl solutions. The effect is believed to be due to the formation of less strongly adsorbed undissociated chloro-complex acids in the case of the HCl solutions.

To better understand the solvent extraction process much effort has been and is being expended at ORNL and elsewhere to measure reaction kinetics of the extraction reactions. To date these efforts have been largely unsuccessful, due to the attempt to apply homogeneous reaction kinetics to two-phase systems. An optical method has been devised for analyzing the concentration profile in the vicinity of a quiescent interface during steady-state diffusion at rather large molecular fluxes. The system under study is the extraction of uranyl ion from water by a TBP solution. Considerable effort has been directed to eliminating leaks in the diffusion cell and testing the optical system/