ABRAC is a three dimensional nuclear thermal depletion program to study the effects of water moderator density changes, resulting from flow variations and boiling, on neutron flux distribution and depletion. The program requires an IBM-704 with a memory of 32.768 words and ten tape units. (auth)

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of …

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 antiproton--proton interaction was studied at three energies, 2.0, 1.25, and 0.98 Bev. Antiprotons produced internally in the Revatron and channeled externally by a system of bending magnets and quadrupoles were selected from background particles by using a gas Cherenkov counter and scintillation counters. At the two lower energies, an electrostatic-magnetic velocity spectrometer was used to reject background particles. A liquidhydrogen target was completely surrounded by scintillation counters so that all charged secondaries from the antiproton--proion interactions could be detected. With the information obtained from these counters, the --p-bar--p total, elastic, inelastic, and charge-exchange cross sections and the angular distribution of the elastic scatterings were obtained at each energy. The total cross section was found to be 80, 89, and 100 mb at 2.0 1.25. and 0.98 Bev. respeclively. The inelastic cross section was about twothirds of ihe total cross section at each energy. It was found that each of the partial cross sections was dropping off slowly with energy. The results were fitted by an optic al-model c alculation. (auth)

Production Test IP-272-A-FP authorizes the installation of up to 100 smooth bore Zircaloy-2 process tubes in C Reactor to demonstrate the feasibility of self-supported fuel elements for production use. An additional 200 zirconium tubes are expected to be delivered by mid-year and con be used to expand the initial demonstration facility. It is the purpose of this document to assess the status of the pilot demonstration program from the B-C Reactor Operation viewpoint.

This report documents the research and testing of an automatic alum feed process at the Hanford reservation. This process generates water of varying qualities which is used to cool fuel elements in nuclear reactors.

A developmental program was conducted to provide and in-pile loop facility for use in evaluating gas-cooledreactor fuel asubassemblies. The program included the design, construction, and installation of a recirculating gas loop which is located in a 6 by 6-in. facility in the aluminum reflector of the ETR. The loop system was designed to recirculate the primary nitrogen coolant at flow rates up to 0.9 lb per sec and pressures up to 200 psia. It will accept fuel subassmeblies up to 36 in. in length and 2.26 ia. in diameter with specimen power generation up to 150 kw. The maximum coolant temperature at the specimen outlet is set at 1500 deg F. The loop system includes the in-reactor section, the machinery, the control system, and the specimen-handling apparatus. Salient features of the re-ertrant system include an aluminum pressure wall in the in-reactor section, static gas insulation between the reactor coolant and the circulating loop gas, and a controllable rate of heat exchange between the specimen inlet- and specimen outlet-gas channels in sections of concentric countedlow piping. The three blowers in the system feature grease-lubricated bearings and water cooling. The complete system was tested out of pile and is now installed in …

An extensive knowledge of the effect on the mechanical properties of metals of prolonged exposure to neutron radiation is considered necessary to properly establish design and operating criteria for in-reactor pressure tubes and test loops. An opportunity to obtain a limited amount of this information on Zircaloy-2 presented itself when, after two years of service, the pressure tubes were replaced in the RE reactor recirculating test facility. Three Zircaloy-2 tubes, with a two-inch inside diameter and 48 feet long, had operated intermittently with prototypical fuel elements at water temperatures up to 250 C (480 F) and pressures up to 1350 psi. During this period, the tubes received an estimated integrated neutron exposure of 1.9 {times} 10{sup 22} nvt. After the tubes were removed from the reactor, metallographic examinations, longitudinal-tensile tests, flattening tests, and burst tests were performed. In this report, the techniques for performing the burst tests are described and the results of the burst tests are compared with the results from tensile tests on coupons cut from corresponding locations along the tube.

As a result of repeated epidemic losses of brazed cathodes at GEXF, the engineering work necessary to determine the best brazing cycle and the controls necessary to insure reproducibility and high yields in the future was carried out by means of EN-297. Included are the procedures used and the tests made to verify the conclusions drawn and the recommendations made.

The differential cross section for charge-exchange scattering of negative pions by hydrogen has been observed at 230, 260, 290, 317, and 371 Mev. The reaction was observed by detecting one gamma ray from the {pi}{sup 0} decay with a scintillation-counter telescope.

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Production Pu nitrate and unfabricated Pu metal during Feb. was below forecast; however FY output is above forecast. Production of UO{sub 3} exceeded commitments; shipments met schedule. Decontamination performance of Purex solvent extraction system was subnormal. Pu nitrate solutions were concentrated. A fire occurred in Purex N Cell during conversion of ion exchange prototype to production facility.

Consolidated Edison type fuel pellets were irradiated and analyzed, to determine the extent of fracturing, particle size of fines produced, and the rate of dissolution in boiling 13M HNO/sub 3/-0.04M NaF -0.1M Al(NO/sub 3/)3. Fused sodium or potassium hydroxide was used to shatter the pellets at 400 deg C or higher. Similar pellets were dissolved in H/sub 3/PO/sub 4/ and fused ammonium bifluoride. An investigation was made of the thermodynamics and limits of flammability of gases expected during the dissolution of sodium-bonded stainlesssteel-clad fuels in aqua regia or sulfuric acid. The amount of hydrogen evolved during Darex dissolution of 304 stain less steel was studied as a function of the fraction of total dissolving time and the total gas evolved. The rate of dissolution of tin in HF containing H/sub 2/O/sub 2/ was> 10 mg/cm/sup 2/- min at 13 to 72 deg C, but decreased> 10 times at 13 deg C when HF was replaced by NH/sub 4/F. A technique was developed for disintegrating and leaching graphite fuels, which yielded a recovery of 99.85% + uranium from fuels containing approximately 5% uranium. The uranium extraction in the Immi hot- cell facility indicated a 0.33% loss in the mixer-settler using the …

The recovery of Th from Blind River ion exchange barrens with di(2- ethylhexyl)phosphoric acid was investigated. The recovery of Tc and Np from fluorination plant residues with tertiary amine was studied. The extraction of Np/sup 4+/ by quaternary ammonium nitrates is reported. A solvent recovery procedure involving successive use of Na/sub 2/CO/sub 3/washing and Al/sub 2/O/ sub 3/- adsorption was demonstrated in laboratory tests as a possible method for the purification and decontamaination of organophosphorus process solvents. The effect of nitrated fractions of Annsco 125-82 on Zr-Nb extractions by TBP was investigated. Treatnnent of TBPAmsco 125-82 solutions with 2 M HNO/sub 3/ at 60 ction prod- C for 1 to 48 hr showed that under these mild conditions the TBP degradation products were more important than those from Amsco as contributors to Zr-Nb extraction and as affecting efficiency of solvent clean-up. The interfacial tensions between benzene solutions of several amine salt and alkyl phosphate extractants and aqueous solutions were examined as functions of the solute concentrations. (For preceding period see CF-59-11-132.) (W.L.H.)

A colorimeter is described that can be used to monitor process solutions continuously for uranyl nitrate or plutonium nitrate concentration. The instrument was tested under plant conditions in the concentration range from 0.1 to 70 grams of uranium per liter and 0.1 to 10 grams of plutonium per liter. The instrument error was plus or minus 1% of the span, but errors of 15 to 20% can be caused by other variables such as acidity and other salts present. (auth)

The nuclear characteristics of the CETR are described. Core operating lifetime, control-rod worth, and powerdensity distribution are discussed in relation to maximizing the core operating life. Other objectives of nuclear design are to minimize the power-density variation and to assure control of the reactor. (J.R.D.)

Methods of relating the bunnable poison concentration in a reactor to the fuel concentration during the life of the core are presented. These methcds correspond to the following ways of using bunnable poison: (1) in discrete lumps, (2) in a homogeneous rnixture with the fuel, and (3) a combination of these. Nuclear data relevant to the U/sup 235/-B/sup 10/ system are presented. (auth)

The special fabrication requirements of the newer columbium alloys, particularly with respect to brazing, stress relieving, weld filler wire, etc., necessitates individual detail and layout drawings for the construction of forced convection corrosion loops from different columbium alloys. Accordingly, it is requested that detail and layout drawings be provided for two forced convection corrosion loops to the specifications outlined.

Experiments were carried out on a Cs/sup 137/ plaque source of 40 x 40 x 1/2 inches. Measurements were made of dose distributions in water and of photon spectral distributions in paraffin. For 9-inch-thick water siab absorbers without air gaps, it was found from the preliminary experiments that the depth dose uniformity can be varied over the range 1.10 to 1.40. The absolute dose rates found were in general agreement with design predictions which indicate irradiator efficiencies in the region of 50% and dose rates in the region of 1 Megarad/hour for high-level indicators. (W.D.M.)

Thorium-uranium alloys were studied with the aim of developing alloys with improved irradiation behavior by control of microstructure. The effect of thorium purity, melting technique, hot and cold working, and heat treatment on microstructure was investigated. The most signifi- . cant microstructural differences occurred as a result of casting technique, The arc-melted alloys exhibited the most nearly ideal structure, that of a homogeneous dispersion of small-diameter uranium particles in a thorium matrix, In addition, the rate of work hardening, recrystallization behavior, density, and hot hardness of thoriumuranium alloys were determined. As uranium content increases, the rate of work hardening increases, The recrystallization temperature of thorium was found to increase by over 100 deg C when uranium is present. Molybdenum, niobium, zirconium, and zirconium in conjunction with niobium were added to thorium- uranium with the aim of increasing irradiation resistance by stabilizing the gamma-uranium phase and/or improving the hightemperature strength of the alloy. It was found that small additions of molybdenum or niobium were effective in stabliizing the gamma-uranium phase, while zirconium was an effective hardener at temperatures up to 600 deg C, Zirconium additions to thorium-uranium alloys were effective in improving the 300 deg C water corrosion resistance of thorium …

The diffusion of hydrogen in zirconium hydride was studied using permeation techniques. The rate of permeation of hydrogen through zirconium hydride disks was measured for small concentration gradients. Data were obtained at 61 to 65 at.% hydrogen and 500 to 750 ction prod- C. The diffusion coefficients were determined by the time-lag method. Ho variation of the diffusion coefficients with hydrogen concentration was observed. The diffusion coefficients can be expressed by D (cm/sup 2/ per sec) = 599 exp (-34,800/RT). (auth)

Losses of fertile and fissile materials during chemical decladding of irradiated prototype Consolidated Edison power-reactor fuel pins by the Sulfex and Darex processes were determined, on a laboratory scale, in all-glass apparatus. For air-fired low-density (-85 per cent of theoretical) fuel cores, minimum losses of uranium, thorium, and plutonium were in the 0.1 to 0.2 per cent range, by either process. These losses increased if the dejacketed cores were allowed to remain in contact with the cladding solution. No selectivity of dissolution of core components was apparent. Comparable losses were obtained with similar unirradiated fuel pins, irradiated core pellets showed a tendency to shatter. When shattered core pellets were present, losses to the cladding solution were excessive. Losses of from 0.5 to 4.5 per cent were observed, depending on the extent of core fragmentation and the time of contact with the cladding solution. No correlation between burnup and extent of shattering was discernible. Core dissolution times were not lengthened by irradiation to the 175 to 300-Mwd/t core level. (auth)

Effective cutoff energies for point l/v absorbers inside of spherical and cylindrical cadmium filters have been calculated for thermal reactor neutrons. The neutron spectrum was assumed to consist of a Maxwellian plus a 1/E component, and the parameters varied were the thickness of filter, the Maxwellian temperature and the Maxwellian to 1/E flux ratio. Because of the sensitivity of the effective cutoff to Maxwellian flux parameters for thin filters it is recommended that filter thicknesses of about 40 mils be used. Forty mil filters show effective cutoffs at about 0.50 to 0.55 ev for temperatures up to about 500 ction prod- A (or about 0.045 ev). Effective cutoff energies for boron filters were also calculated for purposes of comparison. The cutoffs for cylindrical cadmium filters should be applicable to a properly designed experimental facility. (auth)

The effects of the maximum credible accident relative to the design of the containment shell are discussed. The maximum credible accident is defined. The thermal and hydraulic effects of the maximum credible accident on the reactor system were analyzed. The extent to which fuelrod cladding will melt was estimated. The amount of energy released from the reactor system by the escaping steam and water and by a possible chemical reaction was calculated along with the corresponding pressure rise inside the containment shell. The kinds, amounts, and total radioactivity of fission products released to the atmosphere of the containment shell after the core melts were predicted. (M.C.G.)