A specimen of cast thorium-5 wt% plutonium and one of thorium-10 wt% plutonium were irradiated to total atom burnups of 1.9 and 2.6%, respectively, at maximum fuel temperatures of approximately 450 deg C. Both alloys displayed excellent dimensional stability with volume increases of 0.8 and 1.2% per atom per cent burnup, respectively. Three cold-rolled specimens of zirconium-5 wt% plutonium and one cold-rolled specimen of zirconium-7 wt% plutonium were also irradiated. The zirconium- plutonium alloy specimens all showed extremely poor dimensional stability, with anisotropic elongations ranging from approximately 100 to 500%. The irradiation growth coefficients for these specimens ranged from 90 to 210 microinches per inch per atom per cent burnup. The poor dimensional stability of the zirconium-- plutonium alloy specimens is attributed to a highly preferred grain orientation that presumably developed during cold rolling. (auth)

Studies indicated that the use of soluble poisons as a primary criticality control offers economic and other advantages in that it permits the factors of vessel size and shape and solution concentrations to be dictated by considerations other than those of criticality. It is believed that soluble poison criticality control can be made as reliable as other methods of coaditional control if the application is preceded by adequate development work and is monitored by multiple. independent safeguards. The studies included multigroup machine calculations of the required content of poisons in solutions of fissile and fertile material, a compilation of data on the detection, stability, decontamination, and costs of soluble poisons, and an assessment of the possible effects of a nuclear excursion. (auth)

Methods are presented for determining the electric power requirements of a given communications mission in terms of mission and orbit parameters. Analyses were made of possible applications of available and projected space auxiliary power units in these satellites. The satellitc as a communication node is discussed. Example calculations are given. (M.C.G.)

Tests at the SM-1 reactor are reported for the period October 1, 1961, to May 31, 1962. Loading changes were made in SM-1 Core II during the scheduled semiannual shutdowns in October to November 1961 and April to May 1962. Core physics tests include control rod bank calibrations, bank position at several temperature and xenon poison conditions vs core changes and energy release, shutdown neutron source decay and startup channel testing, and critical rod positions for stuck rod configurations. Shielding measurements of gamma radiation in the rod drive pit were made, and dose rates from spent fuel elements as a function of the depth of the water shield were obtained. A lift mechanism for the BF/sub 3/ detector of one startup channel was installed and preliminary testing completed. Water chemistry and radiochemistry tests included a changeover to high pH for the primary coolant, fission product monitoring for iodine, measurement of dose rates on primary system during shutdown, radiochemical analysis of primary water and crud, and change of metal corrosion samples. Buildup of radioactivity in the demineralizer was monitored by radiation surveys and film badge exposures. (auth)

The initial physics tests on the Shield Test Experiment reactor and the precriticality rod-drop test data are presented. (auth)

The JUGGERNAUT is an intermediate-power research reactor designed and constructed as a supporting facility for chemistry and physics research. The design of this reactor is similar to that of the ARGONAUT, and those methods of evaluating the nuclear characteristics of the ARGONAUT which gave good agreement with experimental data were considered applicable to the analysis of the JUGGERNAUT. The analyses for both the JUGGERNAUT and the ARGONAUT were based on a modified two-group theory. The criticality calculations were carried out with the 1BM704 and the two-dimensional PDQ code. Reactivity effects were calculated by hand by means of perturbation techniques, with the real and adjoint fluxes obtained from PDQ calculations. (J.R.D.)

>Thirty papers are included on reactor kinetics with emphasis being placed on reactor safety and design considerations resulting from kinetic work. Information is presented on power excursion programs, reactor transfer function determination and application, reactor instability and thermal-hydraulic problems, and analytical methods in reactor kinetics. The accomplishments in the field and areas needing emphasis are discussed and summarized. Constructive suggestions are made on program direction and information dissemination. Separate abstracts were prepared for each paper. (N.W.R.) lOl2 Data and analytical work on various power excursion tests are summarized and discussed in order to show the present position and understanding of reactor kinetics under accident conditions. The results show that our understanding of plate-type, water-moderated systems of the low power research type seem to be in good shape. On the other hand, information on radiolytic gas formation and transient boiling phenomena is not understood too well. Data are primarily presented on safety experiments with SPERT I and KEWB; however, some information is presented on power excursion tests of Borax I, SPERT III, Triga, Treat, and Godiva. Results show that the problem of predicting the response of reactor systems is on a much firmer basis, even without knowing very much about details. …

The detailed stress analysis of the SM-1 steam generator tube sheet showed it to be safe from strain cycling damage. However, the pressure stresses were greater than the yield strength during the hydrostatic test. The differential between pressure stresses and yield strength indicates that some initial deformation may have taken place in the tube sheet. (auth)

A survey of the pertinent literature was made to ascertain the status of data on U/sup 233/ fission-product yields. The various experimental determinations were evaluated, and the most recent mass-spectrometric results were used as a basis for deriving a set of preferred yields. These yields were compared with values reported in two other recent compilations, and for yields >1%, the three setrs agreed with each other to an average precision of <5%. It was concluded that recent measurements have somewhar improved the reliability of U/sup 233/ fission yield data, but some recommendations for additional experimental work were made (auth)

A copper pulverizer-mixer was designed for homogenizing Molten-Salt Reactor Project (MSRP) fuel. The copper sampling ladle that contains the solidified fuel is placed in the pulverizer-mixer, which is agitated on a mixer mill. The fuel is fractured out of the ladle, pulverized into a homogeneous powder, and transferred to a storage bottle. The homogenized fuel sample is then available for analysis. (auth)

The experimental plan for irradiating BeO pellets in Experiments ORNL 41- 8 and ORNL 41-9 was chosen in accordance with the principles of experimental design. The design is known by statisticians as a 2/sup 5/ factorial experiment confound'' in six replications. Five variables---size, density, grain size, temperature and time--are controlled at two levels to form the basic 2i factorial experiment. The sixth variable, neutron flux, is introduced by confounding on higher-order interactions. An explanation is presented in nontechnical language the means by which the aims of the experimenters and the physical conditions affecting the experiment were utilized in constructing the experimental design. (auth)

The four experimental through-tubes provided in the Experimental Gas Cooled Reactor will extend directly through the core of the reactor and penetrate both the upper and lower pressure vessel heads. Each tube is anchored in an upper head nozzle and the bottom end is allowed to slide in a lower head nozzle. This lower nozzle is basically a T'' section that provides bottom access to the through-tube and a side access for the piping which connects the throughtube to the experimenter's cell. Due to differential thermal expansion of the through- tubes relative to the reactor pressure vessel, vertical movement of the through- tube within the T'' section will be experienced. At the same time a horizontal thrust applied to each tube by thermal expansion of the piping to the experimental cell will result in metalto-metal contact between each tube and the lower T'' section. Tests were conducted on three types of bearing material proposed for use on the through-tubes and T'' sections to minimize galling which can be expected to occur. Stellite No. 12 has been demonstrated to be an adequate bearing material for the intended application. (auth)

A commercial graphite useful in nuclear reactor construction is described. A survey of all currently available sources on chemical and physical properties was made and the information listed. Some data on cost and available sizes are also included. (auth)

Progress on the gas-cooled reactor program is reported. Separate abstracts were prepared for each of the 14 sections. (M.C.G.)

The smear study shows that at least 80% of the floorsmear surveys which were made in a clear area of the Chemical Engineering Building during the course of this study indicated radioactivity equal to or less than 10 d/m/ ft/sup 2/ alpha and equal to or less than 200 d/m/ft/sup 2/ beta-gamma. The smear survey technique is excellent for use in a highhazard area, such as a Pu facility. It is capable of detecting as little as 10 d/m/ft/sup 2/ of alpha contamination with a high degree of confidence. The smear survey is also useful in determining whether the radioactivity on an item is low enough so that it may be removed from an active area. (auth)

To test the fire and explosion resistance of new plutonium metallurgy gloveboxes and to obtain information pertinent to fire control, fire and explosion tests were conducted in one of the gloveboxes. It was found that over l0% oxygen is required for non-metal, and that over 5% oxygen is required for freely burning metal fires. However, metal chips will burn with as little as 1% oxygen if additional heat is furnished. Standard dry chemical, Met-L-X, and carbon dioxide extinguishers were excellent for nonmetal fires. An eutectic salt mixture was excellent for metal fires. (auth)

Rippling phenomena due to heating in fuel plates of SM and PM type reactors are investigated analytically using small deflection theory of plates. Temperature variations across the width of the plate are accounted for. Detailed calculations are conducted for simply supported plates. It is found that within the limitations imposed by small deflection theory that the amplitude of the plate ripples induced by the heating is directly proportional to the initial amplitude. (auth)

None

Run 25 was the final period of power operational of the HRT. The reactor was operated for periods of 62, 8, 52, and 80 hours at 5 Mw with no outward indication of fuel and core and blanket average temperatures of 270 and 230 deg C, respectively. The uranium concentration in the was 1.7 to 2.0 g U/kg D/sub 2/O. Longer periods of operation were prevented by mechanical difficulties, notably with the fuel feed pump. While the reactor was subcritical after the last of the above runs, the upper patch in the core tank wall became dislodged, allowing greater core-to-blanket mixing. The resultant blanket uranium concentration was 2.9 g U/kg D/sub 2/O. The reactor was subsequently operated at April 28, 1961. The experiment was operated at high temperature for a total of 10,866 hours. The system was critical for a total of 8,841 hours and produced 16,295 Mwhours of power. The fuel, heavy water, and some corrosion specimens were recovered, and the reactor was stored in an assembled state. (auth)

A seven- to twenty-fold volume reduction can be obtained from fluidized bed calcination of aqueous aluminum nitrate wastes, depending on the operating conditions employed and their effect on the intra-panticle porosity and absolute density of the calcined alumina. Among the calcining variables, only the bed temperature and the fuel aluminum concentration had a significant effect on the intra-particle porosity of alumina generated during studies conducted primarily in a two-foot-square fluidized bed calciner. A quantitative correlation of the effect of these variables is presented. Alumina with an intra-particle porosity as low as five per cent can be generated by employing a suitable combination of low bed temperature and dilute aluminum feed concentration. Feed sodium concentration and product alpha alumina content were found to have minor effect on intra-particle porosity. Results also show that an inverse relationship exists between the nitrate content of the calcine and the calcination temperature. (auth)

A discussion is presented concerning the preliminary design and hazards evaluation of a H-cooled in-pile experimental loop for operation in the large double-walled through-tube in the Experimental Gas-Cooled Reactor (EGCR) at Oak Ridge. This loop is designed to permit experimentation with full-scale fuel element configurations up to 8 in. OD, at inlet gas temperatures of 600 to 950 deg F at 300 psig, and experimental power levels up to 500 kw. The results of a preliminary hazards evaluation indicate that a loop of this type can be safely operated in the EGCR. The H flammability hazard is controlled by blanketing all H-filled pipes and components with a sufficient quantity of nonreactive gas, such as He or CO/ sup 2/, to produce a noncombustible mixture for all credible H- release situations. (auth)

Fabrication of internal reflector pieces for the SNAP-8 core is described. These reflectors were made of BeO, with and without the addition of Sm/sub 2/O/sub 3/ as a nuclear poison. Because of the high density and dimensional tolerance requirements, the complexity of shapes, and the comparatively modest number of parts to be produced (approximately 1000), the blanks were hot pressed and subsequently machined with diamond wheels and cores. In almost all cases, the specified density of 98% of theoretical was achieved. The low eutectic temperature in the BeO--Sm/sub 2/O/sub 3/ system of about 1420 deg C necessitated special pressing parameters, which were arrived at by experimentation. A rather coarse grit size (80) was used for the diamond wheels, which resulted in very little wear to the wheels, and an extremely low dimensional rejection rate on the blanks. Because of the toxicity of BeO, all equipment was enclosed, and was held under negative pressure. (auth)

Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stability, fission product deposition, gas clean up, activity levels, component and system performance and shielding, and decontamination and maintenance of system hardware. (auth)

Nuclear and thermal analyses were performed to determine the characteristics of the Type 5 core in the SM-1 and SM-1A reactor plants as a function of geometry and composition. The following nuclear properties were investigated: core energy release, maximum midlife reactivity, average fuel burnup fraction, B-10 reactivity coefficient, and power distribution. Thermal parameter surveys determined the effects of channel thickness and power distribution upon the DNBR, nominal and hot channel thermal performance, and fuel plate thermal stress. From the nuclear and thermal analyses, a Type 5 core reference design was selected with fuel plates of 70-mil plate thick ness, 7-mil clad thickness, and 38 wt % UO/sub 2/ in the matrix, having initial core loading o4 108 Kg U/syup 235 and 260 gm B/sup 10/. (auth)