Chemical engineering analysis and stress analysis for design of the 3 megawatt intermediate heat exchanger and steam generator for service with a liquid sodium heat transfer fluid.

Abstract. Results are presented of a theoretical study to calculate the [alpha]/[beta] ratio of filter-collected natural airborne radioactivity. The detection of airborne plutonium contamination by noting increases in this [alpha]/[beta] ratio is discussed. A modification of this approach which provides a more uniform response to alpha airborne contamination is described.

A mathematical analysis is presented concerning the performance of the radial or vane type steam separator. The analyses are concerned with the spread of flow on the vanes because of centrifugal force (which requires that the vane height be greater than the nozzle), and with separation of the vapor from the liquid. (J.R.D.)

From abstract: "A single focusing mass spectrometer has been converted into a double focusing instrument. Relatively high resolution is obtained by decreasing the slit widths and minimizing the ion-optic aberrations. A relatively low ion accelerating voltage (2000 volts) is utilized compared to those normally used for high resolution work. The entire construction was performed at these laboratories. Essentially all of the original equipment was reused in the double focusing system. Although the original source slit was decreased from 10 to 0.4 mils and the source α value was decreased from 0.04 to 0.008 radians, sufficient sensitivity is obtained for quantitative gas analysis with the ordinary electrometer tube amplifiers utilizing 4 x 10[^]10 ohms input grid resistors and 100 per cent inverse feedback. Carbon monoxide-nitrogen mixtures were analyzed quantitatively. The results indicate that direct quantitative analysis can be obtained on mixtures ranging from 99 per cent to 10 per cent carbon monoxide in nitrogen. Analysis of lower percentages may be made possible through indirect reading of the peak should it be necessary. It appears that the instrument should be adaptable to research problems concerning these two gases. The immediate problem posed is the analysis of oxygen and nitrogen in metals. Other …

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Research and development on metallurglcal aspects of pressurized-water systems is summarized. A survey was made of the methods of determining fuel burnup. The mechanisms and kinetics of the loss of boron during heating at 1135 deg C in various dynamic environments were determined. A model was developed to quantitatively characterize the UO/sup 2/ dispersion microstructure of roll-clad fuel plates relative to an ideal'' dispersion. In order to avoid the loss of boron from UO//sub 2/- stainless steel dispersion fuel plates during fabrication, studies were carried out on a refractory glass containing 4 wt.% B/sub 2/O/sub 3/. By using lowsilicon elemental powder, the undesirable reaction between Eu/sub 2/O/ sub 3/ and Si was eliminated; and 13 full-size SM-1 absorbers were fabricated. Work was continued on the borongradient neutron absorber concept. A design was studied for preparing a composite control rod having an upper section made of a boron-gradient dispersion and the lower tip made of Eu/sub 2/O/sub 3/ and stainless steel. Two fuel elements were examined after significant exposure in SM-1. The examination of the miniature boron-iron samples in the final phase of the MTR irradlation test was performed. Twelve miniature test specimens containing 20, 30, or 40 wt % Eu/sub …

Summary report of final hazards for the Boiling Nuclear Superheater (BONUS) Power Station with illustrations.

Report discussing state of the art design and operation of heat transfer and hydraulics equipment and technology in boiling water reactor power plants.

>Information on heat transfer and hydraulics pertinent to the design and operation of boiling water reactor power plants is presented. The following areas of discussion are covered: heat transfer, two-phase density studies, twophase pressure drop, critical heat flux and burnout, boiling stability, calculation procedures for boiling systems, and thermal-hydraulic design procedures for boiling water reactors. (M.C.G.)

Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and training program outline.

Single phase, solid solution, UC-PuC powder was synthesized reproducibly from a mixture of the oxides and carbon. A simple method of monitoring the reaction was developed. UC-PuC-0.1 wt% Ni pellets of 93% theoretical density were made reproducibly, with some samples as high as 95% of theoretical denstty. The pellets contain minor amounts of (U,Pu)/sub 2/C/sub 3/ most of the time. Efforts to make reproducible 95%, single-phase material are under way. Microprobe analysis of UC-0.1 wt% Ni specimens was completed. A second phase apparently promoted by the nickel was shown to be probably U/sub 2/C/sub 3/. Essentially all the nickel was found in solution in the matrix UC. The postirradiation examination of the UC specimens completed to date showed that the UC pellets had the same appearance after irradiation as before irradiation. Their dimensional changes were very slight (1.4% volume increase). Fission gas release was low and within the sensitivity of the initial measurement (3.5% max). The average specimen burnup was 18,500 MW-d/tonne at an average maximum fuel temperature of 760 deg C (1400 deg F). The plutonium facilities continued satisfactory operation at zero contamination level. (auth)

Cs-137 has been chosen as a possible index of the plutonium and uranium in coating waste. This index is a stable ratio as long as the plutonium formation rate (Pu/MD) is stable. Calculations are presented for Purex and Redox. It is recommended that the Cs-137 index method be used for determining nuclear materials content in coating waste at both Purex and Redox; this will result in higher but more accurate statements of Pu, U losses in coating waste.

This report, for January 1962 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.

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 chemical and industrial literature and laboratory work concerned with processing of niobium-containing nuclear fuels are reviewed. 57 references. (auth)

Technical report giving an overview of the use of nuclear energy to produce chemicals including a rough estimate of the economic potential.

Gaseous waste handling systems are compared for direct and indirect boiling water and pressurized water reactors for PL-3 application. Areas that are common to the various concepts are not discussed since they do not enter into a comparison study. The major differences present are in the handling of active gases released to or held in the primary system coolant. These gases which could be present, their possibIe release from the system, and the necessary processing requirements are discussed in detail. Sufficient information is presented to enable one to analyze the methcds required to remove gaseous activity present in other size and type pressurized and boiIing water reactors. (auth)

Activities in a phase of CANDU I refueling machine development are reported which are concerned with obtaining test data for various system subcomponent parts. The basis for selecting those units requiring test was established through engineering evaluation of available information. Results of tests based on these requirements are included in the form of test reports. (J.R.D.)

"This report presents the conceptual design of a 75 MWe prototype Mixed Spectrum Superheater power plant. The scope of the work has emphasized primarily the design, performance, and cost information on the nuclear portion of the plant. The research and development programs required to insure plant feasibility are also present."--Intro.

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 use of continuous solvent extraction was proved quite satisfactory in separating Y/sup 91/ from Pm/sup 147/, using synthetic mixtures of the nuclide pair on a tracer level. The separation was conducted in a compound perforated plate pulse column using a solvent containing 10% HDEHP, di(2 ethylhexyl) phosphoric acid, in toluene. Flow rates to the extraction column were maintained around 1 ml/min of feed, 3 ml/min scrub and 12 ml/min solvent giving one volume change per hour. Pulse frequency was maintained at 50 cycles/min and pulse amplitude at 1 cm. Varying Pm: Y ratio (by activity) in the feed from 0.08 to 135, and using the above conditions, Pm in excess of 99% and Y in excess of 97% purity were obtained. The steady state condition was reached in 12 hours, or l2 volume changes. The approach to steady state was found to be the same for any one component in all runs conducted. However, this approach was much slower than that predicted by calculations reported in the literature for discrete stages. (auth)

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)

Monthly newsletter of the High Plains Underground Water Conservation District No. 1, discussing the field of underground water. Topics include profiles of water conservation research, annual pre-plant soil moisture survey data, annual Winter Water Level measurement data, and information about the latest water conservation tips.

A record is presented of radiation contamination levels observed in the work area during the Core I Seed 2-Seed 3 refueling operations. The radiation levels for Seed 2-3 refueling were somewhat higher thand those for Seed 1-2 refueling in certain areas and are ascribed to operations being carried out under dry conditions. Maximum contamination levels were also higher, but the general contamination level was lower. The activity of canal water was higher and characterization was made on some of the canal water samples. (D.L.C.)