Results of a conceptual design study for a curium powered thermoelectric generator of minimum size and weight which is capable of sustaining hard impact is presented. The generator produces a minimum of 13 watts of d-c power at 3 volts, and weighs 6.2 pounds excluding shielding. (J.R.D.)

This report is about the Advantages of Glass Micro-balloons as Filler Material

Work was continued on the determination of size and distribution of dispersed phase droplets in a pulse column. The droplet behavior and dispersed phase hold-up in Yorkmesh packing was studied with an equilibrated system of methyl isobutyl ketone and water with the ketone phase dispersed. An investigation is being made of the recovery of copper and EDTA from rare-earth ion exchange wastes. The effect of vapor properties on entrainment from bubble cap trays is being investigated. The design, construction, and operation of forced convection loops for circulating liquid metals are reported. The corrosion of 430 stainless steel and 21/2% Cr-l% Mo steel by a liquid lead- bismuth eutectic is reported. The operation of a stainless-steel sodium-cooled liquid-metal condenser is described. The effect of ten additives, including V, Mg, Zr, and Ce in inhibiting the corrosion of stainless steel by lead-bismuth eutectic is being investigated. The current process for preparation of sodium ethyl sulfate is given. Single crystals of Zn, Sn, Pb, and AgCl were grown with a simple, low cost, Bridgman-type crystallizer. Tests were performed at 800 to 1000 deg C on Y, Nb, and Ta loops, circulating Th-Mg and U-Cr eutectics. Work on the development of electromagnetic pumps is …

A listing is presented of documents, films, slides, and those items which were formally produced for utilization by the AEC concerning the SNAP project. (J.R.D.)

Separate abstracts were prepared on 25 sections of this report. A list of publications during the period is included. (C.H.)

Production of Pu nitrate from separations plant during this month was below forecast. UO{sub 3} production, shipments met schedules; so did unfabricated Pu shipments. Purex plant was stopped to replace the final waste concentrator (F-11) (leak). Two Pu-U partition failures were attributed to foreign organic material in the nitric acid; the Pu product was kept within specifications by adding NaF and ANN to ion exchange feed stream. A Np recovery run was started in Redox, and dissolution was started of 12 special 2-ton test batches of normal U fuel elements, irradiated to provide information on Pu formation rates. The damaged B-2 E-metal dissolver was replaced with a conventional dissolver. Test of a new sieve plate cartridge in Recuplex H-1 extraction column was stopped. A new semi-continuous product concentrator-stripper was made to replace Recuplex batch concentrator. Conversion of Purex prototype anion exchange to a manufacturing unit is nearly complete. Design was completed on the new Redox E-metal dissolver. Process feed was introduced into RMC button line and 3 buttons made. Project proposal for NPF reprocessing was revised.

A revision of section 8 of TID-8516, Part I, is presented. The reactor concept which is presented is a change from the former design, a pressurized- pressure vessel-indirect cycle plant to a boiling-pressure tubedirect cycle plant. A description of the plant and a summary of characteristics for the 110- and 325- Mwe systems are given. (J.R.D.)

The reactor design which forms the base for the current economic status of D/sub 2/O-moderated reactors was estimated from developments in several reactor programs. However, since a heavy water-moderated reactor was not operated on natural U fuel at power reactor conditions, considerable improvement from this current status can be foreseen. A summary of improvements is presented concerning the concept which would result solely from operation of succeeding generation plants without a parallel development program, and improvements which would result from the successful completion of the development program as presented. One plant size was used in the evaluation of plant potential, with a 300 Mw/sub e/ nominal rating. The boiling D/sub 2/O-cooled, pressure tube direct cycle plant design was used. The current development program is outlined; this work includes several items leading to the long-range development of the concept. (auth)

An evaluation of 300- and 100-Mwe power plants was conducted using ground rules prescribed by the USAEC for this study. Costs corresponding to two average discharged fuel burnups are: 8.6 mills/kwh (8500 Mwd/ metric ton) and 8.8 mills/kwh (7500 Mwd/metric ton) for the 300-Mw plant. Costs for the 100 Mw plant are 14.7 mills/kwh for an average discharged fuel burnup of 6010 Mwd/metric ton. Estimates of future potential indicate that the 300 Mw/sub 3/ (8500 Mwd/metric ton) plant could produce power for 7.3 mills/kwh in a second generation, full scale plant of the same type. A further reduction to 6.4 mills/kwh should be possible as the result of the recommended ten-year development program. The current development program is adequate for providing the data needed to design and construct a prototype reactor. However, there is no natural U-fueled prototype and no prototype of the chosen reference design scheduled in the U.S. Current technology is sufficiently developed to initiate the design and construction of a pressure tube, boiling D/sub 2/Ocooled, natural UO/sub 2/- fueled reactor prototype plant in the immediate future. This plant would demonstrate the main features of a full scale plant and, in addition. would provide design data which could …

The electromagnetic interaction of an intense relativistic coasting beam with itself, including the effect of a confining nonperfect vacuum tank, or a quiescent rf cavity, is investigated theoretically. It is shown that the resonances that may occur between harmonics of the particle circulation frequencies and the electromagnetic modes of the cavities can lead to a longitudinal instability of the beam. A criterion for stability of the beam against such longitudinal bunching is obtained as a restriction on the shunt impedance of the rf cavity, or the Q of the vacuum tank. This criterion contains the energy spread and intensity of the coasting beam, as well as the parameters of the accelerator. Numerical examples are given which indicate that in general the resonances with the vacuum tank will not cause instabilities, while those with an rf cavity can be prevented from causing instabilities by choosing the shunt impedance at a sufficiently low but still convenient value.

None

A compilation is presented of room temperature and 1200/sup o/F properties of a large number of alloys which might be considered for use in a sodium-cooled reactor. Specific consideration is given to the use of such materials as fuel cladding. The arbitrary basis for property comparison is type 304 stainless steel. The study was made of 10 classes of alloys. These alloys are alloy steels, ferritic stainless steels, austenitic stainless steels, iron and iron-aluminum alloys, precipitation-hardening stainless steels, austenitic superalloy, nickel-base alloys: cobaltbase alloys; copper-base alloys, arid refractory metals and alloys. (W.J.H.)

Tests were conducted to assure proper operation of the control-rod mechanisms of the Shippingport Pressurized Water Reactor under normal operating conditions. (C.J.G.)

Preliminary studies have, in general, indicated the need for modifications and improvements to the reactor last ditch coolants systems in order to provide adequate safety of operation at power levels programmed for the future. These studies have indicated the need for improved reliability as well as increased capacity for the last ditch coolant systems. A design study is being prepared by Reactor Modification Design to define the scope of the modifications required to provide adequate last ditch systems for the older areas. Adequate last ditch cooling will be provided for the 100-K Areas under Project CGI-844 which is currently in progress. The purpose of this document is to set forth the operating conditions and objectives on which the study will be based.

Test equipment to determine the creep properties of Zircaloy-2 in a 750 deg F 500-psi steam atmosphere was designed and constructed. Duplicate tests were performed at stresses to produce failure in approximately 1, 10, 100, and 1000 hr. A comparison between these data and data obtained in vacuum-test equipment showed that a corrosive test atmosphere does not alter ductility or failure times appreciably. (auth)

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Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Currently, Al-Si bonded fuel elements show evidence of spire bond separation, and to a lesser degree, can bond separation following irradiation. Such evidence has aroused concern for the ability of the currently produced Al-Si bonded fuel elements to withstand future reactor operating conditions. Several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary margin of safety for good bond layer integrity. Before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. This report presents the design of a test to more fully evaluate Al-Si bond integrity under anticipated future reactor operating conditions.

The selective removal of halide fission products from an aqueous solution by exchange with the halide in a solid silver halide was studied as the basis for a fuel-element leak detector. The retention of fission-product halides on a silver halide column was investigated as a function of coolant flow rate, halide anion, and column size. Fission prcduct decontamination factors and predicted operating lifetimes were obtained for a number of reactor operating conditions. It is concluded that a sensitive, rapid leak detector for a water- cooled reactor could be constructed from a silver bromide or iodide column monitored by a neutron detector to detect delayed neutrons from the halide fission products. The feasibility of gross gamma monitoring was found to be dependent upon the intensity of the gamma background arising from absorbed fission products on the silver halide column. (auth)

Twenty-four experimental alloys were developed and evaluated as container materials or welding alloys for use with Sulfex and Niflex decladding solutions. Niflex solutions which were more corrosive than Sulfex solutions to most of the experimental alloys, produced severe localized attack on weldments made on vacuum-melted Hastelloy F with the experimental alloys. However, several of the alloys, when self-welded, were not selectively attacked. Some of these showed a substantial improvement in resistance to the decladding solutions. The most promising alloys were based on either 45 wt.% nickel--22 wt.% chromium or 50 wt.% nickel--25 wt.% chromium, with at least 6 wt.% molybdenum, and 1 wt.% titanium, 0.6 wt.% manganese, 0.4 wt. % silicon, 0.02 wt.% carbon, and the balance, iron. The alloy most resistant to both solutions contained 6 wt.% molybdenum and 1 wt.% copper in the 50 wt.% nickel--25 wt.% chromium base. Its corrosion rate of 22 mils per month in Niflex, with no selective attack, was significantly lower than the 105 mils per month recorded for Hastelloy F. Even lower rates would be expected under the less stringent conditions of actual process operation. Indications are that more resistance might be obtained by increasing the chromium and nickel contents. (auth)

A description of a 14-stage electron multiplier for use amplifying the detector signals in a surface ionization mass spectrometer is given. The system can either measure the multiplier anode current or count the anode pulses. Pulse counting may be used for- signal currents in the range from lO/sup -13/ to 10/sup -18/ amp. The range of the normal electrometer circuit is extended by the electron multiplier so that it measures currents from l0/sup -10/ to 10/sup -15/ amp with a fast response. (auth)

Tests were conducted to determine the cumulative and permanent deflection of a beam subjected to two-point loading for (1) ambient temperature- atmospheric environment, and (2) 800 deg F-atmospheric environment, for 8OO psi maximum flexural stresses. Deflection curves are given for five specimens. An analysis of the data indicated that if the core graphite creeps while at EGCR conditions by approximately 1.5 times its creep rate at ambient temperature, the stresses will level off below 1000 psi. Considerable oxidation oc curred for two specimens at 800 deg F. The indicated strain of these specimens may not be representative of AGOT, nuclear grade graphite at 800 deg F in an inert atmosphere. (B.O.G.)

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consisis of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule. (W.D.M.)

The evolution of fission products from UO/sub 2/ during irradiation at high temperatures is of primary interest to the Gas-cooled Reactor Project. Fuel tests consisting of UO/sub 2/ pellets encapsulated in Inconel or stainless steel were irradiated in the LITR, ORR, and ETR. The capsules were pierced in hot cells, and the gases collected in evacuated systems. Fractions of this gas, of suitable activity for counting, were taken and then analyzed by gamma spectrometry. Larger fractions of gas were analyzed by mass spectrometry. Percentage of gas release varied widely, increasing with temperature, impurity content, oxygen-to-uranium ratio of the UO/sub 2/, and decreasing with bulk density. For high density, stoichiometric UO/sub 2/, the gas release was generally less than 3% up to a temperature of about 2800 deg F, about which it was greatly accelerated. Fuel burn-ups of up to 22,000 Mwd/MT were obtained. Maximum measured central fuel temperatures of 3150 deg F were reached. The lower- density nonstoichiometric UO/sub 2/ released greater amounts of fission gas, particularly Kr/sup 85/. (auth)

The fluorination of uranium from dried solids representative of those obtained from the fluid-bed drying of dissolver solutions of high zirconium alloy fuels. e.g.. Dresden and Army Package Power Reactor fuels, was studied under a variety of conditions. Variables investigated were particle size, additives, temperature, time. hydrofluorination, and pyrolysis. Temperatures in excess of 650 deg C were needed to ensure complete removal of uranium from the solid. Pyrolysis with HF of previously fluorinated materials aided in uranium removal from the zirconium solid. (C.J.O.)