Report issued by the Oak Ridge National Laboratory discussing a review of the Radioisotopes Program during 1964. Research and experimental projects conducted during 1964 is presented. This report includes tables, illustrations, and photographs.

Technical report on the activities of the Oak Ridge National Laboratory Biology Division for the report period including a list of 346 publication and lectures and 205 short articles by members of the division on their research and activities.

Report containing a description of the Oak Ridge National Laboratory's High-Flux Isotope Reactor (HFIR) and its operation.

Report documenting the process of decontaminating low-activity fuel waste water via the Scavenging-Precipitation Ion-Exchange process. This includes analyses of the costs of the process as well as a description of the operation of the pilot plant.

Report issued by the Oak Ridge National Laboratory discussing a Health Physics Instrument Manual. Descriptions, characteristics, and applications of different instruments are presented. This report includes tables, illustrations, and photographs.

From introduction: "This is the thirteenth in a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

One of a series of experiments to test the radiation corrosion of Zircaloy-2 and other possible reactor construction materials in UO2SO4 solutions under various radiation intensities, temperatures, solution compositions, and velocities of flow past specimens.

Study of the effects on Zircaloy-2 radiation corrosion of a high concentration of excess H2SO4 in a uranyl sulfate test solution using an in-pile loop experiment.

Report describing ongoing progress made at the Oak Ridge National Laboratory's Thermonuclear 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.

The objectives of this study have been to examine the problems of the pebble-bed reactor concept and to conceive a design of a facility for investigating the feasibility of this type of reactor. The design must provide for adequate leaktightness of the contaminated-gas system and adequate maintenance of contaminated components, the most vital feasibility questions of the concept.

Available knowledge of precision limits in S.S. accountability measurements and/or calculations by reactor and chemical processing groups is surveyed and summarizes. Experience in comparisons of reactor (production and research) calculations versus chemical plant accountability measurements is also reported. A general tentative conclusion is that available precisions (+/- 0.54 to +/- 0.78% ) in chemical plant measurements is also reported. A general tentative conclusion is that available precisions (+/- 1.0 to +/- 11.0%) possible by calculations (nuclear and/or engineering) of power reactor systems; however, with operation and empirical experience (e.g. less than +/-1.0%

Volstility Pilot Plant Mark III Fluorinator will be a double-chamber type vessel, each chamber 2-1/2ft. by 16in. o.d. separated by a 5 in. pipe 15 in. long. ASME flanged and dished heads will be used for the chamber tops and conical sections with a 60º apex angle for the chamber bottoms. A new furnace designed to maintain the complete lower chamber (molten salt + freeboard) above melt temperature will eliminate past experiences of salt solidification on the wall, heads, and in or on the internal process lines. External pipe runs will be sutoresistance heated to allow melting and drain back of salt plugs. The upper chamber serves as a gas de-entrainment and solids precipitation device to retain most of the entrained salt and condensable fluorides in the 100-400°C temperature range.

Results of four experiemental runs in the Fluorox fluidized bed reactor system are reported. The engineering feasibility of UF6 production from UF4 by use of dry air of O2, 2UF4 + O2 = UF6 + UO2F2, in an Inconel fluidized bed reactor at 800-850°C was demonstrated in two experimental tests in which greater than 90% of the theoretical amount of UF6 was collected or measured. Two runs made with crude UF4 (produced from unpurified mill concentrate) as the feed material, showed that UF6 could be produced at 700-725°C but corrosion on Inconel was prohibitive.

At the present time, plants in the 100-500Mv size range are more numerous and carry the greatest portion (over 50%) of the total steam-electric plant load in the U.S. utilities industry. The contribution of plants of over 1,000-Mv capacity is increasing more rapidly than any other size clarification and at present represents about 10% of the total capacity. By 1962 the TVA will have six plants with capacities of over 1,000-Mv. The largest steam-electric plant in the U.S. is the TVA plant at Kingston, Tenn., with a nameplate capacity of 1,440-Mv. Turbine-generator units are also following a trend of ever-increasing size. In present construction, the 150-200 Mv size range for units is the most common and represents the greatest contribution to capacity. Two units of 500-Mv nameplate rating each, the largest in the U.S., are on order by the TVA, and an 800 Mv unit is contemplated.

The design of internally cooled electrical coils for the production of high frequency intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory.

In the use of a reactor as a device to measure neutron cross sections, the pile oscillator is often employed because of its much higher sensitivity compared to that available in static reactivity measurements. The factor limiting the available in static reactivity measurements. The factor limiting the available sensitivity is the generation of random noise, which originates in the statistical nature of the chain reaction. In other words, the nuclear reactions which maintain the chain reaction do not proceed uniformly, but rather at rates which fluctuate in time around some average values. These fluctuations limit the accuracy of measurements in the reactor.

The fire involved approximately 30 liters of isopropyl alcohol as a result of the ignition of about 16-18 liters of isopropyl alcohol and a slurry of thoria and ammonium nitrate in a 20-liter glass carboy. This material was undergoing a volume reduction by boiling in a floor-type metal hood with Plexiglas sides and 3 sliding door on the front of ordinary glass. The alcohol vapor was vents directly into the hood (Fig1.)

Nuclear and radiation hazards have been investigated for all phases of operation to be carried out in the mechanical decladding of SRE fuel elements. The SRE fuel is 2.8% enriched with a maximum burnup of 1,000 Mvd/ton and minimum cooling of 120 days. Each element contains 9 kg of uranium metal (~2 kg U-235) and is made up from 84 slugs (3/4"D x 6'"), clad with (10 mil wall) stainless steel tubing. It is planned to ship the fuel from Santa Susana, California, to Oak Ridge in the modified PAR loop carriers, with a maximum of 10 elements being shipped in each carrier per trip.

The thickness of an oxide film on phosphor bronze helices was determined by first establishing the oxygen content of the helix "as received" and after cleansing with nitric acid. Based on the assumption that the difference between these two values was the oxygen in the film, and that the film consisted entirely of cupric oxide, the thickness of the film was calculated from the density of cupric oxide, weight of the film, and surface area of film. A value of 1080 A was calculated as the thickness by this method.

The 0.045m UO2SO4, 0.036m CuSO4, 0.025 m H2SO4 solution (HRT fuel composition) was chemically stable during 1, 866hr of operation at 280ºC and 1500 psi. The system was pressurized by boiling a 0.4 gpm stream of the fuel in a titanium heat exchanger at 313ºC.

The system Dovex 21K and aqueous uranyl sulfate solutions were studied in the solution compositions range: 0-0.005 mol/liter uranyl sulfate, 0.15 mil/liter total sulfate, 0.02 mol/liter H2SO4 (pH 2).It is shown that UO3SO4 and/or UO2(SO4)2-/2 are sorbed on the resin to form R2UO2(SO4)2. It is surmized that these same species are also involved at high uranyl concentrations and pH 2 it is suggested that U2O5SO4, U2O5(SO4)2-/2, and R2U2O3SO4 species may be involved.

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.