Thirty complete papers and 17 abstracts of papers presented at theFourth Conference on Analytical Chemistry in Nuclear Reactor Technology are given. The abstracts were included for papers to be published elsewhere. Separate abstracts were prepared for the 28 papers. Two were previously abstracted for NSA. (M.C.G.)

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Research facilities, general construction progress, research activities, and administration are discussed and a financial statement is given. Fairly detailed accounts are given of research programs in the fields of physics, accelerator development, instrumentation, applied mathematics, chemistry, nuclear engineering, biology, and medicine. (M.C.G.)

Separate abstracts were prepared on 11 sections of this report. A list is included of staff publications during the period. (C.H.)

Progress made in the development of the Cs/sup 137/ fueled thermoelectric generator for marine power applications is reported. Information is given on thermoelectric generator design, shielding, power conversion system, hazards evaluation, fuel element forming and cladding, fabrication, and heat transfer. Cs/sup 137/ energy calculations are presented along with shielding calculations. The development and fabrication of cesium polyglass for fuel application are included. (N.W.R.)

7 9 7 D 8 8 9 9 7 7 ; 8 < fuel-reprocessing plant using pyrometallurgical procedures is being designed and constructed as part of the Experimental Breeder Reactor No. 2 project. Cable samples sealed with Temporell No. 741 were irradiated to 3.0 x lO/sup 9/ rad. Tests of the absorber-type fume trap were successful. Preparations were continued for high- activitylevel demonstrations of the melt-refining process for EBR-2 core fuel. An attractive and efficient procedure for removing uranium skull material from a melt-refining crucible is to convent it to free-flowing oxide powder by oxidation with a dilute oxygen- argon mixture at 700 to 800 deg C. Two complete runs were made for both demonstrating and pin-pointing difficulties in the drag-out or skullreclamation process. Investigation of the reduction of uranium oxides by liquid magnesium alloys was continued. Two runs were made to demonstrate recovery of plutonium from a magnesium solution by distillation of the magnesium. Operation of a mild steel corrosion loop for circulating a U-Mg- Cd alloy at 550 deg C was terminated after 4800 hr of trouble-free operation. The solubility of titanium in liquid cadmium was found to vary between 0.047 at.% at 449 deg C to 0.15 …

8 7 < 6 ; : : = 8 g developed for recovering fissionable and fertile materials from shortcooled reactor fuels. The second laboratory demonstration of the melt-refining process with highly irradiated EBR- IItype fuel pins was completed. A 392-g charge of U-5% fissium fuel pins irradiated to an estimated burnup of 0.4 total at.% and cooled 28 days was melt refined for three hours at 1400 deg C. Data were not obtained on the behavior of fission products. The effect of N concentration on the nitridation rates of unirradiated U-fissium alloys in Ar-N atmospheres was determined. Experiments on the storage of fuel pins at 350 deg C in Ar atmospheres showed that the presence of 5% N lowered product yields only slightly during subsequent melt-refining operations. Supplementary pouring techniques, such as the use of probes and mashers designed to break crusts over the melts, are moderately effective, but are a less desirable solution to the problem of maintaining high yields than the elimination of contaminants in the Ar atmosphere. A liquid metal process is under development for recovery of the fissionable material contained in melt refining crucible skulls produced in the EBR-II fuel cycle. Information obtained in separate studies …

Proton-deuteron exchange is rapid in boehmite corrosion product formed on pure aluminum in boiling water. In addition, deuterated boehmite films undergo rapid exchange with the humidity of the atmosphere. This explains the previously reported anomaly in the H-D exchange rate for the growing corrosion product on 1100 aluminum. (auh)

Activities in a program to develop techniques of plasma spraying clad plate-type UO/sub 2/ fuel elements are reported. The investigation was also directed toward determining the limitations of the process as applied to fuel element fabrication. UO/sub 2/ powder coatings having densities of 90% theoretical were produced. At conditions required for spraying plates, densities of 86% appear to be practical. The rate and efficiency of UO/sub 2/ coating deposition were also determined for various spraying conditions. Gritblasting was found to provide the best surface for coating adherence. The O/U ratio of the UO/sub 2/ was maintained by spraying in an Ar atmosphere. Zircaloy-2 was found to be the most desirable cladding material. Cladding thicknesses of 0.035 in. are required in distortion-free 2-in.-wide plates. (J.R.D.)

The development of plasma-jet spray-coating techniques for producing clad ceramic fuel plates is discussed. Conditions for spraying fused UO/sub 2/ powder were established by depositing cones on stationary substrates. It was found that the arc-gas flow range within which deposition occurs is very narrow. Coatings were made from --200 +325, --270 + 325, and de-slimed -325 mesh fused UO/ sub 2/ powders. To provide data regarding the economics of the process, deposition rates and efficiencies were determined under various conditions. The effects of powder size, power input, arcgas flow rate, spray distance, traverse rate, power feed rate, powder-gas flow rate, and cover-gas flow rate on deposition efficiency are discussed. Oxygen-to-uranium ratios of coatings made for evaluation of density were determined by gravimetric and volumetric methods. Preparation of the surface without distortion for plasma spraying is discussed. Fixturing and instrumentation methods were designed for measuring substrate and coating temperatures during spraying of typical fuel-element-cladding thickesses of stainless steel and Zircaloy-2. (M.C.G.)

BS>Activities are discussed for work done in the preparation of PuO/sub 2/ using the continuous oxalate process, and continuous coprecipitation studies using a uranium-20 wt% plutonium nitrate feed solution. Characterization studies of the PuO/sub 2/ powders indicated that variations in the processing variables can affect the final product. Measurements with the B. E. T. and Innes apparatus confirmed that the specific surface area of the initial batch of the PuO/sub 2/ powder had increased appreciably during storage. Deltatherm differential thermal analysis apparatus was checked out. Procedures were devised for the determination of plutonium and uranium. Sintering studies were continued for pure PuO/sub 2/, and sintering trials were begun for mechanically mixed and coprecipitated PuO/sub 2/ and UO/sub 2/. Metallographic examinations of PuO/sub 2/ sintered pellets revealed a microstructural feature similar to eutectoid structures in alloys. Mechanical packing experiments were carried out using crushed UO/sub 2/ pellets fired to high density. Plasma torch production of UO/ sub 2/ indicates that excellent spheroidization is attained, but central voids were found in the pellets. Reactor physics studies were completed for the analysis of the potential of plutonium as a fuel in near-thermal converter and straight burner reactors. Plutonium was shown to be promising …

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In this investigation, the effect of an ultrasonic field on the maximum nucleate heat flux (burnout) that can be sustained by boiling water in a flow system will be determined. The water will flow in the direction of sound propagation within an annular flow channel bounded on the outside by a glass tube and on the inside by a 0.25-in. OD heating element. The design of the test section and flow system is given. (D.L.C.)

The continuing analysis of the VBWR core resulted in refinements in the calculations for reactivity in voids, flux leakage and resonance escape probability. The Zircaloy cladding for 25 fuel assemblies was received and passed inspection. Preliminary measurements of VBWR flux oscillations, used to develop instrumentation and data interpretation techniques, showed random normally-distributed oscillations with a predominant frequency of 0.5 to 1.0 cycles/second. A model for analog computer simulation of a reactor as a feedback control system was adapted to VBWR. Equations for the hydraulics model and preliminary results from use of the model are presented. Irradiation of the Fuel Cycle stainless steel clad assemblies reached 412 MWD/T with specific powers of 28 kw/kg (average) and 52 kw/kg (peak) during January. Visual examination of the fuel after this irradiation indicated that it is in good condition. The VBWR was shut down during February and March for replacement of all in-core components made of 17-4 pH stainless steel with 304 stainless steel. The details of the first eight special fuel assemblies were determined and materials were ordered. The effects of steam quality, mass flow rate, and rod diameter on burnout heat flux are shown. The burnout heat flux varied inversely with mass …

The Fuel Cycle Development Program is a basic development program for boiling and other water technology. It covers the areas of oxide fuel fabrication. irradiation. and examination; the physics of water-moderated reactore; and boiling-water heat transfer and stability. Schedules for the fuel- cycle program were examined. and it was concluded that portions of the Task A program should be conducted during the period May to Dec. 1959 in order to keep costs of the work as low as possible and to allow initiation of the fuel-cycle program at the earliest possible date after the Vallecitos BWR was returned to service. The basis for the scheduling of the work is discussed. and a chronological summary describing the content of the work is given. Technical progress is outlined and details are summarized. Subsequent reports issued monthly and quarterly will summarize the progress of the prognam. (W.D.M.)

Coatings on the fuel element surface and coatings on individual fuel particles are being investigated for retaining fission products in fuel for the Pebble Bed Reacter. Ten fuel element specimens with surface coatings were subjected to varying amounts of high level irradiation. Cracks or pinholes were found in 6 of the coatings. Evidence indicated that the graphite matrix contributed to most of the failures. Release factors of the order of 10/sup -9/ for the Si--SiC coating under high-level irradiation existed for a period of about one month, however. The pyrolytic carbon coating showed some promise as a fission product barrier in several neutron activation tests. Fourteen batches of UO/sub 2/ particles were coated with A1/sub 2/O/sub 3/ by the vapor deposition process. Tests showed that this coating is an excellent barrier to fission products. In pyrolytic carbon-coated UC/sub 2/ particles there is no temperature limitation due to reaction between the particle coating and the graphite matrix. (M.C.G.)

Emphasis was shifted in the Pebble Bed Reactor (PBR) Fuel Element Development Program from coatings on the sphere surface to coatings on individual fuel particles as the major deterrent to fission-product leakage. In a highlevel irradiation test, cracks developed in the coatings of specimens coated with pyrolytic carbon and siliconized silicon carbide. In another high-level irradiation test, a graphiie sphere fueled with Al/sub 2/O/sub 3/-coated UO/sub 2/ particles is showing excellent fission-product retention. The leakage factors for long-lived volatile flssion products such as Kr/sup 85m/, Kr/sup 87/, Kr/sup 88/, Xe/sup 135/ are ranging from 10/sup -9/ to 10/sup -6/. If this degree of fission-product retention is maintained in a large power reactor, it would result in essentially a "clean" primary loop. A simple crack in a fuelelement surface coating will permit the release of all of the volatile fission products in that specimen except those retained by the fuel particles. In view of the failures in surface-coated specimens tested to date, it appears to be a difficult task to ensure coating integrity in a large number of specimens because of their low thickness-to-diameter ratio and exposure to external loads on the fuel element. The test of a single specimen fueled …

Fabrication of alumina-coated UO/sub 2/ and pyrolytic carboncoated UC/ sub 2/ particles was studied. Some reaction was noted between alumina and graphite at 2500 deg F. For UC/sub 2/ particles coated with carbon at 2000 deg F, the coatings were found to crack at temperaturss above 2000 deg F, whereas 2450 deg F deposition gave fewer failures at 3600 deg F, more rapid deposition- ratss, and absence of excess soot formation. A Pebble Bed Reactor fuel element consisting of a 1.5-in. graphite sphere fueled with alumina-coated UO/sub 2/ particles was irradiated at 1400 deg F to a burrup of 3.3 at.% U/sup 235/. Up to 2.5 at.% burnup, the fission product leakage factors (rate of release/rate of production) ranged between 10/sup -9/ and 10/sup -5/ for 10 isotopes and were ascribed to trace amounts of uranium contamination outside the particle coatings. In the latter part of the quarter, ths leakage factor for Xe/sup 133/ rose to lO/ sup -3/ while several other shorter lived fission products increased to a smaller extent, indicating the beginning of diffusion through the coatings due to radiation damage of the alumina. Carbon-coated UC/sub 2/ particles coated at 2000 deg F were found to have good …

Progress is reported in 26 papers on radioactive waste disposal, ecologicah research, radiation physics and dosimetry, internal dosimetry, and health physics technology. Twenty-five separate abstracts were prepared. One paper was previously abstracted for NSA. (M.C.G.)

The effect of various dipping times, temperatures, and diffusion treatments was determined for tantalum sheet dipped in aluminum alloy baths. Both aluminide and beryllide coatings were produced that will withstand oxidation for 10 hours at 2500 deg F, isothermally and cyclic. Aluminide coatings were obtained on a stressed niobium alloy that meets the same test conditions. (auth)

Eperimental Breeder Reactor I. The fully ribbed and rigid Mark III loading of EBR-I was found to be govenned by feedback processes which guarantee safe and stable operation under normal operating conditions and to give a large radial contribution to the power coefficient. Nonlinearities in the power coefficient were investigated and found to be no problem. If the stabilizing ribs are removed from the fuel rode, a strong positive effect appears which is associated with the inward bowing of fuel rods. The prompt positive coefficient obseved in Mark II is discussed from the standpoint of Mark III tests. A 800-Mwh irradiation run was made on a number of samples, and some bric cladding failures are reported. Data are given for the dimensional changes in EBR-I, Mark III fuel rods used for a total of 2,682 Mwh operating time; the fuel rods usually increased in diameter and decreased in length, and some bowing was obseved. The growth and temperature profiles of the fuel rode are compared, and the effects of radial restraint on the rod growth are discussed. The EBR-I, Mark FV core design is then discussed. The fuel rod will incorporate four plutonium-10 at.% aluminum fuel slugs with two depleted …

High-vacuum, high-temperature equipment for rapid, accurate determinations of liquid-metal solution compositions is described. The solubility of uranium in bismuth was redetermined over the temperature range 300 to 725 deg C with improved precision and temperature calibration. Anomalous solubility effects in uranium-bismuthgraphite systems which could affect the operation of reactors containing graphite are described. (auth)

Progress and results from the conceptual design, economic evaluations, and research and development work performed as part of the Nuclear Superheat Project are reported. Developments in conceptual design and program evaluation, fuel technology, materials development, experimental physics, coolant chemistry, heat transfer, mechanical development, SADE, and mixed spectrum superheat design are discussed. (M.C.G.)

Tests performed on the Shippingport PWR to determine the magnitude and sources of primary coolant water leakage from the reactor coolant system are described. Two of the pressurizer relief valves are found to account for 80% of the total leaksge of 56 gal/hr from six pressure rehief and reactor relief valves. Other valves are found to have insignificant leakage. The total primary coolant water leakage is found to be 26 gal/hr. (auth)