The original goal of this investigation was to compare the thermodynamic characteristics of the gases in and behind the shock fronts in gases at initial pressures in the millimeter range and to compare these characteristics in the geometries of single and double discharges. The shock fronts were not visible, so it was not possible, at these pressures, to get visual data from the shock front itself. The parameters giving the properties of the gases were faces. Measurements made with an image converter camera (which is still in the development stage) agree well with these made with a photomultiplier tube. Differences are observed between the front velocities in the cases studied. These are of the order of 3 to 15 per cent. Considering the nature of the shot to shot fluctuations in the discharges and the inductance variation between the single and double discharges represent a physical difference. The mathematical treatment which says that two equal strength colliding with a wall behaves, has not been shown to be inadequate by this investigation. It was hoped that a stronger confirmation could be fien to the theory, but the accuracy of the data does not warrant it.

The corrosion of INOR-8 and Inconel dissolver components used in the fluoride volatility process for the dissolution of zirconium with anhydrous HF in molten salts was investigated. Ten dissolution runs were made using simulated subassemblies formed from Zircoloy-2. The dissolver and components were made from INOR-8. Both the dissolver vessel and draft tube were examined several times during the series of runs. The solids which formed at different areas in the system were also analyzed. The results showed that the corrosion of the INOR-8 dissolver was greatest at the salt-off gas interface and at the HF gas inlet. Almost all of the corrosion took place during run 10 when no zirconium was present. Portions of the dissolver were cleaned after run 10 and sent to BMI for evaluation. The results of the Battelle metallographic examinations of the portions are included along with several photographs. The results with Inconel tubes in the copper-lined hydro-fluorinator confirmed the observations that the liquid gas interface areas were the most susceptible to attack.

Since the inauguration of the Off-Site Animal Investigation project n 1957, there has been one annual report rendered as of 30 June 1958. this is the second annual report dated as of 30 June 1959. The objectives of the project have been unchanged during the past year. These are (1) to enhance the Nevada Test Site/off-site rancher relationship through an active investigation project in their interests, and (2) to provide further information as to the status of the off-site animals in their environment, with special emphasis on their radioactive isotope uptake from fall-out. isotope uptake of the animals is being emphasized as well as the gross and microscopic histopathological examinations. Two year's existence in an area of radio-contamination where a full fission spectrum of isotopes can be found, where radiation background reading range from 1/2 mr per hour to 1 = per hour, has produced no noticeable effect on the animals.

The purpose of the test was to determine the neutron flux level in a BF3 counter well during the second performance of DL-S-225. The measured activity of the irradiation brass bolt was due entirely to the radioactive decay of Zn65. The thermal neutron flux in the BF3 counter well during the second 1000 hour run was calculated to be 2.5 x 10(9) neutrons/cm2 sec.

Results are summarized on theoretical considerations of the excited states of the Ca isotopes, experimental studies of the level structure of Ca42 and Ca44, studies of the production of circularly polarized bremsstrahlung by beta rays, the Moller scattering spectrometer, and the Moller scattering coincidence experiment.

The purpose of the test was to determine the radiation levels inside the concrete enclosures but outside the reactor plant containers after shutdown following plant power operation. Radiation levels at the survey points in the 1-AC and 10BD Boiler Chambers Enclosures and in the Reactor Container Enclosure indicated that no significant radiation hazards were present approximately 25 minutes after all rods had been inserted. The radiation levels approximately 4 minutes after shutdown at the survey points in the Auxiliary Chamber Enclosure indicated that several points were above background, the highest test level being obtained in contact with the East Auxiliary Chamber container drain pipe.

The purpose of the test was to check the extraction crane rails in the area of the reactor pit for level and parallelism. The west extraction crane rail exceeded the allowed tolerance of 1/32 inch at only one location. The elevation of the north bumper was out of tolerance by 1/64 inch. The east extraction crane rail was consistently lower than the west rail by as much as 8/64 inch. The east rail was parallel with the west rail within allowable tolerances over the length tested except at one location where the transit was located, approximately 41 feet from the north bumper.

The purpose of the test is to determine the reason for the decreases in the liquid levels of the reference chambers serving the reactor plant remote liquid level indicators. The remote gages and the local liquid level gages agreed closely for the Reactor Plant Component Cooling Water Expansion Tank and the Coolant Discharge and Vent System flash tank and blow-off tank. There appeared to be no loss of water from the D/P cell reference chambers for these two systems over the period of the test. There was no definite indication of leakage from the Valve Operating System reference chamber, however, the differences between remote indicator readings and the sight glass readings are attributed to sticking of the ball check valve on the upper sight glass, inaccuracy of the readings and instrument error. It is recommended that he lower shut-off valve on the upper sight glass be cleaned and reopened prior to reading the sight glass to ensure that the ball check valve is not stuck in the closed position. No lead age was found in any of the level indicating systems.

In conjunction with Gas-Cooled Reactor heat-transfer studies, local and mean heat-transfer factors are predicted from the heat transfer-mass transfer analogy using subliming naphthalene in air. Experimentation was conducted on 1-in.-dis septafoil rods in a 4-in. -dis flow channel with rod center-to center spacings of 1.10, 1.25, and 1.40 in. at a Reynolds modulus of approximately 60,000. Ratios of local mass transfer to mean mass transfer for a given rod vary as much as from 0.7 to 1.3 (outer rod, 1.10-in. spacings). Mean values of the mass-transfer factor are, in general, above that predicted by the correlation j-0.023 NRe^-0.2; as much as 46% got the outer rod t 1.25-in. spacing. The data indicate that for maximum mass transfer and minimum variation of the mass-transfer factor, an optimum rod spacing exists; the best observes is at 1.40-in.

This is the sixth in a series of interim reports describing various phases of the study of fuel element rupture kinetics and mechanisms. Previous reports issued are: No. 1- Experimental Methods and Procedures, HW-61378, No. 2- Coextruded Rod Elements with Pinhole Defects, HW-61379, No. 3 - Mechanism of the Uranium-Water Reaction, HW - 61799, No. 4 Coextruded Tube Elements with Pinhole Defects and Various Annular Spacings, HW- 62348, No. 5- Comparison of In-Reactor and Ex-Reactor Ruptures, HW-62766. This work is being done in cooperation with the Fuel Element Design Operation of the Hanford Laboratories Operation. J. W. Goffard has been particularly helpful in supplying samples and useful discussions of the results.

The majority of two-phase flow problems involving equations of state are solved by use of point-wise utilization steam table values. In this manner, problems involving the use of the various flow equations of continuity, momentum and energy are generally forced into iterative solutions. Considerable effort towards the development of an analytical expression for the state equation seems indicated so as to simplify the analysis of two-phase problems, particularly apparent in the analysis of systems undergoing phase transformation as demonstrated by the significant difference between simple theory and experimental critical flow determinations. The assumption of homogeneous, equilibrium mixture is indicated as a first attack upon the problem.

Design Test Request PR-20 Calandria Characteristics, outlined the need for experimental data concerning the performance of the calandria under transient conditions. Test data was required to confirm that the moderator dump system would drop the level the required 24 inches in less than one second. The original calandria dump chamber design was modified until the criteria was met. This information is recorded in HW-58333, Interim Report, Design Test PR-20, Calandria Characteristics, which lists the drop for the first 24 inches only.

Through the years, considerable effort has been expended in studies of the explosive reactions sometimes observed in the dissolution of uranium-zirconium alloys in nitric acid. It has been shown (1) that such reactions result from the rapid oxidation of finely dived solids released by the preferential dissolution of the metallic matrix. The explosive portion of such solids has been identifies (1) as an intermetallic compound with the approximate composition UZr2. This compound, referred to as the epsilon phase in previous work, has more recently (2) been termed the delta phase. The latter designation will be employed here.

Two water quality control labs are being provided for the the NPR. One, a "cold" lab, is located in the183 Building adjacent to the control room. Its primary purpose is to provide facilities for quality control of the output of the filter plant and the demineralizer plant. The other, a "hot" lab, is located in the 190 Building. Its primary purpose is to provide facilities for the quality control of the primary and secondary coolants, and the moderator coolant.

A study of the nuclear safety in the storage and transportation of PRTR fuel elements has been made. This study was based on 7-rod clusters of plutonium-aluminum allow fuel elements containing 1.8 per cent Pu by weight. Each cluster is 7 feet 4 inches in length and contains 270 grams plutonium. Drawings of the "New Fuel Storage Pit" (H-3-11030) have been reviewed for nuclear safety. Nuclear safety criteria for the design of a lead shielded fuel transfer cask as well as criteria for the storage of these fuel elements outside the facilities mentioned in the above drawings have also been reviewed. For water moderated systems, a homogeneous model of plutonium, aluminum, and water was used t evaluate the critical parameters. These results should be conservative. At the conclusion of an experimental program to determine criticality parameters of PU-Al alloys in light water, a theoretical approach will be developed to calculate such criticality parameters.

Water treatment systems for preparing and maintaining high purity water in out-of-reactor or in-reactor test oops are becoming increasingly important. In out0of-reactor experiments the presence of ionic impurities in the water has a marked influence on film formation and corrosion rates. It is therefore , imperative that these impurities be maintained at the lower practical concentration.

Ceramic insulated, stainless steel sheathed thermocouples have been used to monitor temperatures of encapsulated uranium specimens, both in-reactor and out-of-reactor. No operational difficulties are encountered at low temperatures, but at a temperature of 700 C or greater, a eutectic is formed between uranium and iron. This reaction destroys protective sheath and results in thermocouple failure. A typical example of the phenomenon has been reported by J.W. Geffard of the Fuels Development Operation. Hanford Laboratories. Tantalum was suggested as a barrier between these metals and an evaluation of this system was made at 500 C.

Abstract: Fission product measurements were made on the SM-1 primary coolant. The airborne activity observed during the sampling of the primary system was identified. An analysis was made on the primary coolant for alpha activity and on the secondary water for fission production iodine.

Abstract: Critical experiment studies were performed, varying the parameters U235, B10 and metal to water ratio, in the SM-2 7 x 7 core configuration with 38 stationary elements and seven control rods of the SM-1 (APPR-1) type. An experimental mock-up of the SM-1 was assembled using the basic SM-2 fuel plates. Excellent agreement between the SM-1 boron loading, determined by chemical analysis, and the SM-1 mock-up boron loading, for equivalent bank positions, was noted. Several SM-2 mock-ups, cold clean and midlife, were assembled and studied with regard to reflector effects, flow divider effects, relative control rod array worths, critical rod configurations, and relative power distributions. The results of these experiments indicate as satisfactory a U235 loading of 36.4 Kg and a B10 loading of 63.4 grams for the SM-2. Attention is drawn to numerous power peaks present in the active core. The open seven control rod array has a slight reactivity advantage over the closed seven array and consequent minor disadvantage with respect to "stuck rod" criteria.

This quarterly progress report discusses the ongoing work within the Solid State Division at the Oak Ridge National Laboratory for the period ending August 10, 1952. Projects discussed include radiation metallurgy, engineering properties, fused salts, crystal physics, and solid state reactions.

This report is the final in a series of preliminary reports that follow the studies of scavenging systems related to radioactive fallout. The project consisted of two phases: preliminary experiments to relate the sizes of particles in air to specific radioisotopes, and preliminary laboratory studies of scavenging of particles by liquid drops, including studies of sticking probability and effects of Brownian motion and water vapor diffusion.

The following report describes the workings and development of the nuclear ramjet propulsion systems.

Report discussing theories of irreversible thermodynamic processes. "The purpose of this review is to collect the presently available experimental data for a variety of quite different irreversible phenomena and to show that this evidence does indeed verify the Onsager Reciprocal Relations."

The following document was done under the auspices of the U.S. Atomic Energy Commission with the intention of analyzing air core cryogenic magnet coils and its usage in fusion research and high energy nuclear physics applications.