A study was made of the adaptation of the HAPO anion exchange process to the reprocessing of Fast Ceramic Reactor (FCR) fuel using the Idaho Small Plant Concept. It is shown that the anion exchange flowsheet meets the reprocessing objectives of the FCR case and can be adequately accommodated in the Small Plant Concept. Capacities of up to 1550 Md(e) are feasible in the Small Plant and unit reprocessing costs range from 0.14 to 0.28 mills/kwh depending on the number of reactors to be processed. (auth)

ABS> The removal rates of iron, nickel, and chromium from synthetic stainless steel waste solutions during electrolysis over a mercury cathode were studied. The loading capacity of the mercury for the stainless steel metals was estimated on the basis of laboratory experiments to be about two% by weight. The laboratory data indicated that, at an electrode potential of --1.80 voits vs S.C.E., 85 ampere-hours per liter of waste removed essentially all of the stainless steel ions from a sulfuric acid solution containing 0.13M metal ions at 35 deg C. (auth)

The distorted-wave theory of direct nuclear reactions is presented in a unified manner, in which the effects of assuming various reaction mechanisms and nuclear models appear only in certain radial form factors. The zero-range approximation is used, and spin-orbit coupling is neglected in the distorted waves. Formulas are given for transition amplitudes, cross sections, and polarizations. A description is given of the IBM-704 computer code SALLY that is based on these formulas. (auth)

The purpose of this document is to present the numerical results of a production test of the last ditch water backup system performed at B and C reactors on December 15, 1961. The main purpose of the production test was to determine more accurately the capacity of the export system for supplying flow to a dual reactor area under several simulated emergency conditions where BPA power to all areas has been interrupted and the steam supply to one old area has been lost. In addition, the flow capacity of the high tanks to,the reactor was tested alone and in parallel with the export system, and the action of the surge suppressors at all areas was recorded when cycling of the surge suppressors was deliberately induced.

BS> The potential advantages of neutron radiography as an inspection method are discussed along with a historical review and discussions of neutron sources and detectors. The results of the current investigation of neutron-image detectors are discussed in regard to photographic speed, relative neutron-gamma response, and resolution comparisons. Two neutron-image detecting methods are discussed. In one, the direct-exposure method, both the converter screens and the film are exposed to the neutron beam together. The other method, the transfer method, makes use of a radioactive, image-carrying screen, which is transferred to photographic film after the neutron exposure is completed. The direct-exposure method results in increased speed, but has the disadvantages that the film also responds to any gamma radiation in the imaging beam and that, in most cases, improved image resolution can be obtained with the transfer method. Reference is given to several application possibilities. (auth)

The generation of secondary missiles by blast waves was investigated in Operation Plumbbob for three nuclear detonations with estimated yields of 11, 38, and 44.5 kt. A trapping technique was used to determine the impact velocities for 17,524 missiles (stones, glass fragments, spheres, and military debris or steel fragments) which occurred in open areas, houses, and an underground shelter with an open entryway. The equivalent ideal-wave peak overpressures computed from measured blast data for the open-area stations varied from 3.8 to 21 psi. Two houses and an underground shelter were located where the overpressures were 3.8 and 65 psi, respectively. The effect of hill-and-dale terrain on the production of missiles was investigated on one of the shots. Precursor effects were noted on two of the shots at stations near Ground Zero. Missile velocities measured at all stations except the underground shelter were compared with those computed by use of a model based on an ideal blast wave. An analytical procedure was presented by which translational velocities of nmn can be estimated using the measured velocities of spheres and stones. Total distances of displacement were measured for 145 stones that weighed up to 20 kg and for 1528 fragments from a …

Recovery of Th (and U) from Granitic Rock. Recovery of Th by acid leaching ten addltlonal granite samples (36 to 82 ppm Th) from the Conway formation in N. H. ranged from about 50 to 85%, and averaged about 70%, Study of the effect of grind size on the recovery of Th from Conway and Plkes Peak granites showed no significant differences in the range minus 20 to minus 200 mesh. The Th concentration in a sized Conway granite sample was found to be much greater in the fine than in the coarse fractions, whereas Pikes Peak granite showed only slight Th enrichment in the finer fractions. U recoveries in acid leaching of four different granite samples were not improved by adding an oxidant. Collection and Analysis of Granite Samples. A field survey of the Conway granite formations in N. H. was made. Preliminary analysis of the data indicates that the accessible surface of the Conway granite averages at least 40 ppm Th. Collection and Analysis of Lateritic Soils. The Th concentration ranged 5 to 16 ppm in twenty-two samples of sub-lateritic soil from Miss., Ala., Ga., and Va. Final Cycle Pu Recovery by Amine Extraction. In continued batch countercurrent …

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.)

The geologic map of the Paiute Ridge (Papoose Lake SW) 7 1/2-minute quadrangle shows an area of about 4 square miles on the east side of Yucca Flat that is underlain by basaltic rocks. These rocks are estimated to be as much as 1,000 feet in places and occur as a capping on a small mesa and as lopoliths, igneous masses that in cross section are saucer-shaped. This area was further studied to determine the geologic features of the basaltic rocks in order to evaluate the potential use of this area for underground test sites in media other than tuff, alluvium, and granite, particularly in media of high density. The results of this work are set forth in this report. There are two possible test sites in basaltic rock in the area. Site 1 (within Nevada State coordinates E 709,500 and 710,500 ft and N 854,000 and 856,000 ft) is suitable for a shallow test. Site 2 (bounded by Nevada State coordinates E 707,000 and E 712,000 ft and N 848,000 and N 853,000 ft) could be used for a deep confined test in basaltic rock.

A criticality excursion occurred at the Oak Ridge Critical Experiments Laboratory on November l0, l961 as enriched uranium metal, neutron reflected and moderated by hydrogen, was being assembled. It is estimated that the energy yield was between 10/sup 15/ and 10/sup 16/ fissions. There was no personnel exposure or property damage. Fission product contamination, both airborne and contained in the metal, decayed sufficiently overnight to allow unhindered continuation of the experiment. The excursion was caused by a too rapid approach of the two sections of uranium constituting the experiment. (auth)

Progress Development of a high temperature, vaporfilled thermionic converter for application with a nuclear reactor for space-vehicle electrical power generation is reported. Problems associated with the design and opera tion of a thermionic converter employing a UC-ZrC emitter, a cesium plasma for space charge neutralization, and a high-temperature collector are described. Emitter fabrication techniques are also described. A test cell employing a cylindrical UC-ZrC emitter, which was pressure bonded to a tantalum sleeve, and a low- temperature copper collector, was fabricated and operated for 400 hours to provide experimental data. The emitter was operated at temperatures of the order of 2000 deg C while the collector temperature was maintained at 200 to 300 deg C. A conceptual design for a thermionic power reactor incorporating the thermionic converter under development is also studied. It was concluded that a thermionic fuel element would be about 20 inches long and 0.68 inch in diameter and would incorporate 1O thermionic cells. The load voltage per fuel element would be about 14.5 volts and two elements would be connected in parallel (electrically) to provide an output of 29 volts. The over-all design would provide an electrical power level of approximately one megawatt. (auth)

The increasing incidence of leaking process tubes, and the approaching end of the useful life of process tubes in the C and K Reactors, have focused attention upon the various sources of aluminum process tube leaks. Further, the replacement of large numbers of process tubes with new ones, also of aluminum, will require continued attention to these sources to make efficient use of the new tubes. One of the sources of process tube leaks is the corrosion that attacks the interior surface of the process tube. The factors influencing the extent of this corrosion attack are varied and complex, and in recent years, the corrosion service conditions have become increasingly more severe. Each of the factors involved in determining the rate of corrosion attack has thus become individually more important, and the need to understand the inter-relationships among them has increased. It is the purpose of this report to discuss the technical factors contributing to the internal corrosion of the process tubes, to review the way some of these factors have varied in the past, to examine the means available for evaluating the extent to which corrosion has damaged the tube walls, to comment upon the ways in which knowledge …

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The efficiency of multiple traversal targets is defined as the probability that a proton dies by making a nuclear collision in the target rather than by hitting the limit of the synchrotron aperture. The efficiencies of Be, Al, Cu, and Pb targets are shown for 15 and 30-Bev protons in the Brooknaven AGS. Beryllium was found to be the most efficient. (M.C.G.)

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In a prevlous paper, some inequalities occurring in the one-veloclty theory of neutron transport with lsotropic scattering were derived. Some of the previous results are generalized to the case of linearly anisotropic scattering. (auth)

The status of the analytic portion of the KEWB program at the time of completion of the spherical core experiments is summarized. Three computer programs were developed for use in this analytic effort. The first reassembles and smooths three decades of reactor power data read separately from oscillogram records of reactor excursions. It then computes the logarithmic derivative of the power, energy release, fuel solution temperature, and temperature compensated reactivity. The second program utilizes the space-independent neutron kinetics equations with any number of delayed neutron groups to determine the reactivity in the reactor from the power and its derivative. The third program solves the space-independent kinetics equations for the neutron flux from an input reactivity or initial period. Up to 50 reactivity feedback equations includirg delayed neutrons are provided for in this program. A mathematical model of the reactor investigated extensively was one containing six delayed neutron groups, conventional treatment of temperature reactivity compensation, and void compensation of reactivity induced by radiolytic gas void growth proportional to the product of reactor power and energy release. Partial mathematical solutions to the kinetic equations were derived for reactivity feedback proportional to prompt temperature and void growth according to the product of power …

The interests of several departments at Hanford are involved in the planning, execution and evaluation of the results of the KER loop testing effort in support of the NPR fuel program. The varied interests and activities of the participating groups must be well-integrated if effective use of our limited testing capability is to be made. The purpose of this report is to help achieve this integration by summarizing the current thinking on the goals of the NPR fuel testing program and by presenting the current loop schedule.

The effect of polarizing mngnetic field intensity on the nuclear resonant absorption was studied by varying the field strength at a 1-mc Co/sup 57/ source from 0 to 1000 gauss while keeping the absorber between the poles of a magnet having a fixed field of 800 gauss. The rates of resonance absorption change with field intensity were greatest in the region of 300 to 1000 gauss, and the% nuclear resonant absorption for 1000-gauss fields was 8.5 and 26% for perpendicular and parallel fields, respectively, as compared with 15% for no fields. Other absorption measurements for Co/sup 57/ sources are also reported. Calculations on the use of nuclear resonant absorption to measure gravitational fields and altitudes were made which indicates that this application is not promising. (D.L.C.)

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Tests of Cambridge absolute filters, Model Sl-071, specified for use in the GCR-ORR Loop No. 2 as full-flow, primary coolant fiiters were completed. kD.O.P/ (dioctylphthalate) efficiency tests were performed on three filters in the as-received condition, on two filters following canning and thermal cycling, and on one of the canned fiIters following bsking out. None of the three units met the design criteria of 99.97% efficiency for removal of 0.3 micron particles in the as-received condition. The postthermal cycle efficiencies of the canned fiIters were slightly higher than their respective as-received efficiencies. At the corapletion of testing, the two fiiters canned for installation in the reactor facility had measured efficiencies of 99.855% and 99.93%. These values were judged acceptable for the intended application/su The thermal cycling of the two canned filters and the subsequent baking out of one of these units demonstrated that a limited amount of off-gas products would be given off/su Pressure drop tests were performed on the canned fiiters with instrument air (ambient temperature, atmospheric pressure) over a flow rate range of 150 to 530 lb/hr. Curves of pressure drop across each fiIter versus Reynolds number were plotted for air and He. (auth)

EQUIPOISE-3 is an IBM-7090 FORTRAN programmed code for the solution of two-group, two-dimensional, neutron diffusion equations. A maximum of 2l00 mesh points may be used, and the code will solve problems in either rectangular or cylindrical geometry. Logarithmic derivative boundary conditions are allowed, and removal of neutrons from both groups is permitted. Adjoint fluxes with the associated fluxadjoint flux regional integrals may be calculated automatically if desired. A constant buckling, group-dependent buckling, or region-dependent buckling may be specified for rectangular geometry. This program is intended to fill the need for a rapid two-dimensional calculation suitable for survey calculations. During the iterative part of the computations, all operations are carried out in the core memory. The magnetic tape memory is used only for input, output, and program storage. The running time for a 1000-point problem requiring 100 iterations would be about 3 min.(auth)

A fuel jacket failure that occurred in May 1961 in the Type 304 stainless steel clad fuel element exposed in the Vallecitos Boiling Water Reactor superheated steam loop (SADE) was attributed to chloride stress corrosion cracking. In order to better understand the failure, a test program was carried out to try to reproduce the rapid stress corrosion attack in the simulated superheat reactor environment of the CL-1 superheat facility. The methods of corrosion testing under heat transfer conditions reported previously were modified: to apply a longitudinal stress on the test sheaths to produce a 0.1 per cent elongation in 1000 hours; to increase the chloride content of the moisture carryover with the steam by increasing the chloride in the recirculating water to 1.5 ppm; and to expose the solids deposits to various metal temperatures. After 1000 hours of exposure, no significant attack was noted on the test sheaths. The test procedures were further altered to simulate the significant amount of SADE fuel element exposure to saturated steam at varying temperatures with little to no superheat being generated. A 776-hour total exposure was carried out with the test conditions cycled several times. The entrance heater (calculated metal temperature during normal operation …