The thermophysical properties of Armco iron such as thermal conductivity, electrical resistivity, and Seebeck coefficient have been extensively investigated and reviewed up to 1000 degrees C. Few investigations of such properties have been made on high purity iron. If such a study is made using the same apparatus to determine the properties of two purity levels of iron, then several significant intercomparisons can be made which add meaning to data on a single material. The systemic errors for a single apparatus are the same, therefore comparison of a property of two similar materials is more significant. A comparison of the property changes with temperature and purity can show the effects of impurities on the mechanisms contributing to a property and allows prediction of the properties of iron as a function of purity. For these reasons a study was initiated on the high-purity iron for comparison to Armco iron.

The experimental details, mathematical models, and typical data for a rapid comparative method for thermal conductivity measurements are presented. The method consists of measuring the temperature change of a small silver sphere after it is brought in contact with a small disk-shaped specimen which was initially at ta higher temperature. This temperature change was calibrated in the range of 50 to 400 degrees C by making measurements on samples of know thermal conductivity. The accuracy of this technique was shown to be between than +-10% with a reproducibility of at least +-2.5%. Using known transport mechanisms for heat conduction in solids and the temperature dependency of the electrical conductivity, a means to judiciously extrapolate thermal conductivity data obtained between 50 and 400 degree C to high temperature is presented.

Grade CGB graphite is a nuclear graphite which is basically an extruded petroleum coke bonded with coal tar pitch. No carbon blacks are used and the low-permeation graphite is finished through a series of impregnations and heat treatments with a final heat treatment of all components to 2800 degrees C. A listing of the results obtained is given in Table 1. The results at 51 degrees C are considered questionable. There was a slight contamination of the 90% Pt 10% Rh-Pt thermocouples at 910 degrees C but it was not sufficient to doubt the validity of the 910 degrees C results. However, the results obtained at 1015 degrees C should be disregarded because of severe thermocouple instabilities. In addition, the electrical resistance of the core heater at 603 degrees C indicated the thermocouples had a -10 to -15 degree error which is sufficient justification to disregard the 605 degrees C data.

This technical report describes development work done on method of particle separation by the Biology Division of the Oak Ridge National Laboratory and the Oak Ridge Gaseous Diffusion Plant during the period January 1 to June 30, 1963, under the Joint National Institute for Health-Atomic Energy Commission Zonal Centrifuge Development Program. The central effort has been to develop zonal centrifuge systems for the separation of cells and sub-cellular particles, including viruses, and bio-colloids, including proteins and nucleic acids.

Report containing a series of reports from members of the Oak Ridge National Laboratory's Neutron Physics30 Division.

Report issued by the Oak Ridge National Laboratory discussing progress and work conducted by the Chemistry Division.

Sanderson & Porter have carried out a series of studies over the last four years which indicate that the pebble-bed reactor way be an attractive way to obtain low-cost power. At the request of the Atomic Energy Commission, two design studies have been carried out on this concept at the Oak Ridge National Laboratory. The first of these a preliminary design of a 10-Mw(t) reactor experiment, the PRRE, was initiated September 10, and a report on the study was issued November 1960. The second phase of the work, a conceptual design study of a 330-Mw (e) central station, was initiated November 1, and is the subject of this report.

A carrier containing 138.99 liters of solution, uranium concentration 202.04 g/liter with an isotopic concentration of 97.3% U-233, was prepared for shipment. The total uranium was 28,062 +/- 60 g (95% confidence level) and the U-233, 27,305 +/- 66 g (95% confidence level).

Report documenting ongoing research and developments at the Oak Ridge National Laboratory's Gas-Cooled Reactor Project.

A portable shield has been designed, developed, fabricated and shop tested to provide the HRT with a facility for direct dry maintenance operations. It provides temporary replacement for any one of the lower roof plugs and should permit many operations to be performed without flooding the reactor cell with water.

The effect of adding small quantities of fuel or poison to the HRT has been estimated using perturbation theory. The results have been reduced to a single relation and a set of graphs which make the estimation of added reactivity relatively simple. The perturbation theory results are compared with multigroup results and reasonable agreement is demonstrated; however, there is some question concerning the prompt neutron lifetime.

Temperatures are given at which two liquid phases form in a synthetic homogeneous reactor fuel solution and its concentrates. The data show a two-liquid-phase boundary temperature of 332°C for the Particular HRT Fuel composition and a flat minimum temperature of 305°C for the initial solution concentrated between 300 and 329°C are presented to indicate solution stability in this temperature region. Some related comments on current HRT operations are given.

Several experiments have been performed at ORNL with light water solutions of uranyl nitrate (highly enriched in either U^233 or U^235) in an essentially bare sphere 27 inches in diameter. This report presents the results of several calculations with elementary bare reactor theory and a discussion of the observed discrepancies between the calculated and experimental results. If the observed critical concentration is used in the calculations, the calculated effective multiplication constant is less than unity' thus a higher critical concentration would be predicted than is actually observed.

The use of mercury as a solvent in the recovery of uranium from spent fuels is of the interest at Oak Ridge National Laboratory. The vapor pressure of mercury is lowered by increased concentration of uranium. By dew-point measurements, the vapor pressure at 175°C was found to very between 2 and 8mm of mercury, and at 375°C, between 300 and 1100 mm of mercury, depending upon composition as described below. Plots of the log of mercury vapor pressure vs. the reciprocal of absolute temperature gave a family of straight lines. Each line corresponded to one of the composition: UHg2, UHg3, UHg4, and a saturated solution of UHg4 in Hg. No Mutual solubility of the intermetallics was indicated.

In the operation of nuclear reactors, nuclear fuel reprocessing plants and in-pile experiments, special provision must be made for disposal of gaseous fission products to prevents contamination of the atmosphere to an unacceptable degree. A disposal process is described in which the noble gas fission products, krypton and xenon, are delayed relative to the sweep gas by physical adsorption as they pass through an adsorbent such as activated charcoal. A theoretical plate analysis, and has been verified experimentally. The retention time for a gas present in trace concentration is proportional to the amount of charcoal in the adsorber bed and to the adsorption coefficient which is evaluated experimentally for a particular combination of materials and conditions. The retention time is inversely proportional to the volume flow rate if the sweep gas.

The technique of high-frequency titrimetry has been applied to the determination of thorium, uranium, sulfate, and free acid. In Part I of this report, the reproducibility of the method for the titration of standard solutions which contained 50mg of thorium in the absence of interferences is established. The coefficient of variation of the method, under these conditions, was found to be less than one per cent. In Part II, the effect of uranium on the high-frequency titration of thorium, as well as the application of the method to actual samples, is discussed. Uranium in a ratio of 5 to 1 to thorium can be tolerated. When the method is applied to the analysis of representative samples, the coefficient of variation is one per cent.

After two batches (~ 340 kg) of fluoride salt from the ARE were reprocessed, pilot plant operations were terminated because of a leak through which an estimated 780 g of uranium (as UF6) escaped. Of the 21 kg of highly enriched uranium in the feed, 93.12% was collected as UF6 product, 0.13% represented measured losses, and 3.72% was unaccounted for (leak). An additional 3.03% was reclaimed from NaF beds and equipment washes. The product met both chemical purity and activity specifications for product level UF6. Decontamination from fission products was essentially complete. A gross gamma D.F. was apparently limited by the low activity of the feed salt.

Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types, have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of rare gases, iodine, bromine, cesium, and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50% of the rare gases, 33% of the iodine, 9% of the cesium, and traces of strontium. After 25% burn-up, the cesium value increased to about 60%. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2% of the iodine, 10% of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15% burn-up, when melted at 1850 C, released up to 95% of the rare gases, 90% of …

A series of two-group calculations has been made on the Oracle for the purpose of obtaining critical-mass and flux distribution data for various ORR core configurations. The 3G3R code of Bate, Einstein, and Kinney was used, together with the RSP code developed by Nelson. This made it possible to obtain results for the three-dimensional case. The results, which are presented graphically, are intended to serve as a guide for the design of experiments until such time as actual measurements are available. The calculations were performed for the "clean cold" case, and it should be realized that the presence in the core of experiments and of fission products built up during operation will materially alter the flux patterns found. It is believed that the critical-mass data are accurate to within 10%. Within the fuel region it is believed that the thermal-flux patterns are the also accurate to this degree. Comparison of the results with MTR critical experiments, however, indicates that the thermal flux in the reflector in the vicinity of the fuel-reflector interface may have been underestimated by a factor of as much as 1.3. It should also be recalled that in a two-group calculation the "fast flux" is often a …

The continuous dissolution of 304 stainless steel and stainless steel-UO2 alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver product to remove chloride ion was studied in a 4-in. diameter Pyrex bubble-cap column containing 12 single bubble cap plates. Continuous dissolution rates and dissolver product stainless steel loading were correlated with aqua regia feed composition, acid feed rate and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained …

An annual report is given on the operation and costs of waste-disposal facilities at ORNL laboratories and operating buildings in the Bethel Valley area. The operations of the hot-chemical and metal-waste systems, the process-waste system, and the radioactive-gas-disposal system which utilized the 250-ft stack located in the Radioisotope area are discussed. The miscellaneous operations which include the SS (source and special nuclear) material control, SS material recovery, off-shift service for research divisions, water demineralization plant operations, and hydrogen liquefaction are included. However, the disposal of cooling water from LITR, off-gases from the Hot Pilot Plant, and the ORNL Graphite Reactor building are not covered by the report.

The OCUSOL-A program (ORNL-CF-57-6-4) for Univac is a modification of the Eyewash (ORNL-1925) multi-group, multi-region reactor code. The group=energy-lethargy-temperature relationship are given in Table A. The element code numbers are given in Table B. The cross sections now on the sigma-tape are given in tables in the Appendix numbered with the element code number. This technical report explains the bases for choosing the cross sections.

An investigation of the causes of observed explosive reaction of zirconium-coated uranium dioxide on dissolution in nitric acid was conducted. It was concluded that such a reaction is to be expected. Possible but unconfirmed methods of alleviating the problem are suggested.

A general program has been written to plot curves on the Oracle curve plotter. A description of the code and complete instructions for preparation of input tapes and operation of the code are given. The code tape is available from the Mathematics Panel or from the author.