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The design, performance, and cost information on the nuclear portion of the Mixed Spectrum Superheater power plant are emphasized. The research and development programs required to ensure plant feasibility are also presented. The nuclear steam supply system, reactor auxiliary systems, radiation control systems, control and instrumentation, special test instrumentation, plant operation and maintenance, steam cycle, turbine plant, general service systems, preliminary safeguards considerations, expansion of plant power output to 150 Mw(e), and MSSR critical experiment are described. (M.C.G.)

An analysis is presented of the deflecting mode in an iris-loaded synchrotron waveguide which is not based on the concept of TE and TM modes as generating fields. The solutions employed are derived from the transverse components of the electric and magnetic Hentzian factors. Their use avoids the singnlarities in the field expressions for phase velocities equal to that of light, which must occur with TE and TM modes. The fundamental properties of the hybrid solutions are exposed. The lowest mode of the hybrid group is the desired deflecting mode. The exact relation between frequency, phase velocity, and geometry of the structure can be derived for the deflecting mode in the form of an infinite determinant. Transverse and hybrid solutions are tabulated. (J.R.D.)

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Progress made in the development of the Experimental Beryllium Oxide Reactor (EBOR) is reported. The objective of the EBOR program is to develop a gas-cooled, beryllium oxide-moderated reactor which can be used in conjunction with a closed-cycle gas turbine or a steam cycle for a small land-based or a maritime power plant. Progress is reported on reactor development, reactor physics, and materials development. (N.W.R.)

S>The compatibility of type 304L stainless steel in intimate contact with graphite is being studied as a function of temperature and contact pressure. This study is an outgrowth of materials compatibility problems in present and advanced gas-cooled reactors, where structural members in direct contact with graphite provide the possibility of both carburization and self-welding. Initial studies were concerned with surface reactions in the absence of gaseous contaminants under a vacuum of 10/sup -6/ mm Hg at 540 to 705 deg C. Stainless steel specimens are pretreated to provide three surface conditions: H/sub 2/- fired, preoxidized, and Cu-plated. Surface contact pressures ranged from 0 to 10,000 psi. Test results are presented which establish the lower temperature limit for significant diffusion between graphite and stainless steel at approximately 60O deg C. Above this temperature, diffusion between untreated or H2-fired stainless steel surfaces was found to effect complete bonding of the two materials at contact pressures as low as 500 psi. Bonding was effectively prevented by the presence of either an oxide film or a Cu plate at temperatures up to 700 deg C. Where bonding occurred, diffusion rates measured for C in stainless steel were comparable with those reported for stainless steel …

The specification contains information concerning the material, process, and product requirements to be met by the fabricator. (J.R.D.)

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)

An addition is presented for the IBM-7090 program MURGATROYD to produce a rough graph of reactor power versus time. A sample of output is included for the case given as an example. (J.R.D.)

The effect of modifying the dislocation structure by room-temperature prestraining on the subsequent yielding behavior at 77 deg K is being studied. Results on tantalum single crystals indicated that a considerable decrease in the early yield stress for a given strain is effected by prestraining at room temperature. Twinning was observed in the tantalum at 77 deg K and normal rates of strain. It is shown that the potential barrler to dislocation motion in crystalline solids can be measured in relatively pure bcc transition metals such as Nb, Ta, Mo, and W. Work is being carried out to extend the method of differential calorimetry to measurements of stored-energy-release spectra in deformed bcc metals. Measurements for copper are discussed. (M.C.G.)

Tests were carried out on the self-actuated reactor relief valves and the self-actuated pressurizer steam relief valves in order to insure their reliable operation. The valves operated properly, and the final set pressures were all within the specified limits. (D.L.C.)

A test to obtain data for use in determining the reactivlty wohh of Na in the Hallam reactor core is described. The test is designed to obtain information on the dry temperature coefficient of reactivity and to train operators. An evaluation of results is included. (J.R.D.)

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On September 19, 1961, the No. 3 horizontal control rod at KW reactor was removed after an apparent failure of the wall separating the boron carbide powder from the coolant water. The instrument indications that were interpreted as a rod failure include: high radiation alarms in the exhaust air, unexplained gain in reactivity, increasing coolant outlet temperature on No. 3 HCR, and high radiation readings of the outlet water from No. 3 HCR. The possible reactor safety aspects of such a failure made it necessary to obtain a thorough examination of the rod and inner coolant tube. A complete borescope examination of the rod and partial visual examination of the inner coolant tube have recently been completed. This document is intended to summarize the inspection results, discuss the safety and costs aspects of a horizontal rod failure, and suggest courses of action for the remaining rods at KE and KW reactors.

The General Electric Company, as prime contractor to the AEC at Hanford, is proposing to modify the lattice characteristics of the 8 3/8-inch lattice reactors for the purposes of improving the conversion ratio of these reactors. The proposed overbore modification of the reactors would remove the existing aluminum process tubes, enlarge the diameters of the graphite channels by about one-half inch, insert smooth-bore Zircaloy-2 process tubes and refuel the reactor with larger size, self-supported fuel elements. The overbore fuel will remain the internally-and-externally-cooled cylindrical type, but the weight per foot will be about twice that of the present fuel element. The removal of the inlet and outlet piping connections which would be required in the overboring process will permit the replacement of the existing fittings with ones of improved design. Furthermore, new orifices and venturis which are compatible with the hydraulic characteristics of the overbore tube and fuel geometry and the pumping system will be installed. No basic changes are proposed in the pumping system though the reactor flaw rate may be increased 5--10 percent by changes in hydraulic characteristics depending on the water plant flow capacity.

The New Production Reactor, as compared to existing Hanford reactors, employs new and unique design concepts. To properly evaluate these design concepts and their effects on reactor operations, nuclear safety, and reactor life, a comprehensive testing program is planned; this program, with its objectives and restrictions, is discussed in this report. It has been developed along the same line as the C and K Reactors test, programs, and is expected to require a total time of 6--8 weeks of round-the-clock testing. This estimate includes fuel loading time, but does not include time allotments for engineering acceptance tests prior to power operation nor does it include any of the time necessary for engineering and physics tests during the extensive power ascension program. The main body of this report is presented in three parts. The first section describes startup hazards and restrictions, reactor and component safety provisions prior to loading, and the itemized listing of quantities to be measured. The second includes preliminary material and plant conditions and a brief description of the individual tests. The third section (the Appendix), written in procedure language, comprises a rather detailed description of each individual test on a tentative basis; final test details and procedures, …

These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

One of the early candidate fuel elements for N Reactor use was the coextruded, Zircaloy-2-clad seven-rod cluster. As part of the program of evaluating the seven-rod cluster geometry, three-rod cluster fuel elements, two and three feet long, were irradiated. These long cluster fuel elements were irradiated to determine the distortion (or sag) which might occur at the center of the unsupported length during irradiation. Two three-rod clusters made up of 0.630 inch diameter rods, containing natural uranium cores were irradiated in KER Loop 3. The rods of one cluster were three feet long; the rods of the other were two feet long. The three-feet long rods were supported at their ends and at their midlengths, the two-feet long rods only at their ends. During the irradiation, the maximum core temperature was 435 C. The fuel elements were discharged from the loop after they had reached an exposure of 1400 MWD/T. Following the discharge, the fuel elements were visually examined in the KE view pit. No sag was observed in any of the rods. The test demonstrated that two- and three-feet long rod cluster fuel elements can be irradiated without appreciable sag occurring in the rods.

One of the early candidate fuel elements for the N Reactor was the seven-rod cluster fuel element. An objective of the program to determine the suitability of the seven-rod cluster fuel element for N Reactor use was to evaluate the irradiation performance of coextruded, Zircaloy-2-clad, seven-rod cluster fuel elements over a range of exposures from low exposures to high exposures. This report describes the irradiation testing of an enriched seven-rod cluster fuel element which was irradiated to 520 MWD/r.

River temperature traverse data within and just the plant boundary are quite limited. For this report, traverse temperatures were calculated by first averaging the depth measurements vertically at each traverse point, and then weighing the average for the cross section by the depth at each point and the span between points. Calculated {Delta} T`s were determined by dividing the total reactor power level above the traverse line by the river flow, with the appropriate factor to obtain the correct units. No attempt has been made to apply corrections for atmospheric effects or the diurnal cycle.

This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company.