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The properties of electrons in a gas and an external electric field, e.g., agitation velocity, free path, energy lost per collision, and collision cross sections, were determined for electrons of energy less than 2 ev in a large number of gases. The computations were made from previously measured values for Townsend's energy factor or the ratio of electron agitation energy to molecule thermal energy. Results are presented for hydrogen, nitrogen, carbon dioxide, methane, ethylene, cyclopropane, and argon. (D.L.C.)

7 7 6 4 : 9 8 9 6 5 : 9 aluminum nitrate waste was calcined in the close-coupled continuous evaporator-pot calciner unit. Waste oxides from TBP-25 waste were incorporated into glassy materials after the addition of phosphate and borate fluxing agents. Melts formed at 850 to 950 deg C were glassy solids when cooled and had densities from 2.41 to 2.47 g/ml. Waste volurne reduction factors were from 7.6 to 9.3. Low-Level Waste Treatment. A demonstration run was completed in the 60 liters/hr scavenging-ion exchange pilot plant on ORNL low-activity waste. Decontamination factors were satisfactory after 1500 resin bed voluumes of waste had been treated, i. e, Sr> 1000, TRE 100, Cs> 100, and sufficiently high for other activities that the waste activity was reduced to <10% of MPC. The physical properties of vermiculite, clinoptilolite, and rock phosphate were found to be suitable for second-stage treatment of process waste. Engineering, Econommc, and Hazards Evaluation. A cost study of the conversion of high-level solutions to solids by pot calcimation was completed. Aging had a negligible effect on costs for processing in a given vessel size. The lowest cost was 0.87 x 10/sup-2/ mill/kwh/sub 3/ for processing acidic Purex and …

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<><DSN>16:009503<ABS>Included are summaries of studies of: DCX-1 operation and performance; DCX-2 design and engineering; plasma physics; vacuum arcs; ion production, acceleration, and injection; plasma theory and computations; magnetics; and vacuum system techniques. Separate abstracts were prepared for the eight sections. (B.O.G.) 9504(Faye unscannable abstract)

A dry criticality test was carried out to determine the minimum critical mass of the HNPF Core without sodium. A subcritical calibration of the central control rod was performed and the relative reactivity worths of the inner ring of six control rods were determined. The extrapolated critical loading for the various plots after each incremental fuel loading with all rods out is shown. A tabulation is presented of multiplication data taken throughout the dry critical test. In order to find the relative integral reactivity worth of the central control rod in the dry critical loading, subcrltical multiplication data were obtained. The test completion criteria as stated in the test procedure were met. (M.C.G.)

BS>Intercalibration data for various temperature sensing elements in the reactor coolant system are obtained and compared with calculated source activity levels and observed channel levels. The readings of 20 of the 62 core thermocouples were either less than 400 deg F, which was off scale on the recorder, or were plus or minus 20 deg F from the calibrating resistance thermometer temperatures. The remainder of the readings were within plus or minus 3% of the temperatures of the calibrating thermometers which were used as a standard of comparison. The main coolant loop resistance thermometers all read within plus or minus 0.2% of the calibrating thermometers. The data from the nuclear source range instrumentation indicated a level ranging from less than 1.7 cps to 31 cps; Channel A indicating the highest. The recorded source activity levels were generally higher than the calculated source activity levels as they should be. There appeared to be no correlation between coolant temperature and source log level source range indications. (N.W.R.)

Progress in nuclear safety research and development is reported. Topics covered include: reactivity effects of fuel displacements in a pool-type reactor, release of fission products on out-of-pile melting of reactor fuels, release of fission products on in-pile melting of reactor fuels, fission product transport evaluations, characterization and control of accident-released fission products, nuclear safety pilot plant, preparation of reactor containment hardbook, and radiochemical plant safety studies. (M.C.G.)

An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)

The authors report a study of the reaction K{sup -} + p {yields} {Lambda}{sup 0} + {pi}{sup +} + {pi}{sup -} produced by 1.15 Gev/c K{sup -} mesons and observed in the Lawrence Radiation Laboratory's 15-in. hydrogen bubble chamber. A preliminary report of these results was presented at the 1960 Rochester Conference. The beam was purified by two velocity spectrometers. A hyperon/observed during the run and the preliminary cross sections for various K{sup -} reactions at 1.15 Bev/c have been reported previously. Reaction (1) was the first one selected for detailed study, because it appeared to take place with relatively large probability and because the event, a 2-prong interaction accompanied by a V, was easily identified. In a volume of the chamber sufficiently restricted so that the scanning efficiency was near 100%, 255 such events were found. These events were measured, and the track data supplied to a computer which tested each event for goodness of fit to various kinematic hypotheses. The possible reactions, the distribution of events, and the corresponding cross sections are given in Table I. An event was placed in a given category of Table I if the {chi}{sup 2} probability for the other hypotheses was < …

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Using the Berkeley polarized-proton target, we have measured the polarization parameter P({theta}) for proton-proton (p-p) scattering. The measurements were obtained at beam kinetic energies of 0.330, 0.680, and 0.740 GeV at the 184-in. synchrocyclotron and 1.70, 2.85, 3.50, 4.00, 5.05, and 6.15 GeV at the Bevatron. The angular regions measured were from 20 to 100 degrees center of mass; the square of the four-momentum transfer ranged from 0.1 to 0.8 (GeV/c){sup 2}.

Extensive precipitation was observed in S8ER cladding following irradiation. Electron microprobe analysis was proposed as a means to identify these precipitates. Subsequently, the electron microprobe was successfully used for qualitative analysis of unirradiated and irradiated S8ER cladding. The unirradiated, thermally aged control specimen analyzed showed two major precipitates. One, a randomly distributed intergranular phase, contained Ti, Mo, Fe, Cr and Ni and a second, a randomly distributed intragranular phase, contained Si, Mo, Fe, Cr and Ni. The irradiated specimen analyzed showed the same precipitates. The phase containing Ti, Mo, Fe, Cr and Ni was also randomly distributed intergranularly. The phase containing Si, Mo, Fe, Cr and Ni was in the grain boundaries, and preferentially segregated to the outer and inner diameter.

The object of this test was to obtain data concerning the reactivity of the horizontal rods in the E-N fuel loading at H Reactor. An accurate calibration of the control rods could then be used to provide a more accurate measurement of the other reactivity variables associated with the E-N loading. A comprehensive rod calibration test was performed at H Pile in 1957. Results of the present test, needed because of the ``blacker`` E-N fuel in the pile, have been normalized to the 1957 test results and extrapolated to a full-system calibration. Two short tests were carried out for this purpose: (1) a scram transient test, and (2) a modified rising period calibration test.

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The objective of this production test was to evaluate the effects of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining. Test details are provided.

Irradiation of a third thermocouple train with a thermocouple element of the same design as used on the previous two trains, but with different heater elements, is authorized to an exposure no greater than 1000 MWD/T. A second decontamination of KER-1 using the same procedures as for the first decontamination is also authorized.

This document lists the physics factors for the K Reactors following the transition to the KV fuel element geometry and zirconium tubes. Each new parameter with the zirconium tube lattice has been calculated relative to the factors used with aluminum tubes and the KIV fuel elements. The purpose of this document is to provide working values for plant assistance use during the transition to the zirconium lattice. In some cases, where there are large uncertainties in the absolute values, the conservative end of the range has been provided for present operational use in safety and control administration. Refinement and publication of ``best`` values for the zirconium lattice based on the extensive experimental and calculational studies are included in future Reactor Physics Unit programs.

Tube life estimates for the base case and the five study cases are given for 6.6 and 7.0 pH. The fuel sizes and operating conditions for each case are summarized. Information on the mixer position that gives maximum tube life for each case is shown. A tube life distribution was calculated for the base case using the outlet temperature distribution for 1197 central zone tubes at B reactor. Average tube life for the 1197 tubes war estimated to be 3.0 years (100 percent TOE) for operation at 6.6 pH, and the outlet temperature corresponding to average tube life vas l07{degree}C. Similar numbers could be calculated for the study cases if estimates of the outlet temperature distributions were available.

A testing program designed to provide the necessary data for evaluation of the emergency flow backup adequacy was conducted at the K reactors on February 16, 1963. Three transient or trip-out tests and 13 steady-state tests were performed under authorization of PT IP-573-I. This report contains the numerical results of the pressure and flow data that were obtained during these tests.

This study was developed to provide a correlation for the evaluation of the pile C{sub RI}/Tube C{sub RI} ratio (alpha value) for each of the Hanford reactors.

The scope design for Birch recovery at Redox was presented in September 12, 1960. Final plutonium purification and decontamination by anion exchange in a pulsed-bed contactor under development by Hanford Laboratories was proposed. Potential advantages of low cost and improved operability were anticipated from the new contactor; when these were not demonstrated in the development work in time to meet project schedules, it was decided to revert to the Higgins-type contactor for which considerable Purex operating experience can be drawn upon for design bases. This scope revision presents the necessary changes to cover the change of contactors.

This report explains a mathematical algorithm designed to complete contour descriptions of weapon components which have been only partially specified by free-form data.

Information concerning the pressure-flow relations in reactor process tubes at very low heat generation rates is of value for determining flow requirements during shutdown conditions. Since such information is quite difficult to calculate when boiling occurs, a brief laboratory investigation was conducted to obtain experimental data. The data obtained for BDF geometry are recorded in this report for distribution to those persons having an interest in the application of the results to reactor operation.