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AN 80 MEGAWATT AQUEOUS HOMOGENEOUS BURNER REACTOR. Reactor Design and Feasibility Problem

Description: An 80 Mw aqueous homogeneous burner reactor suitable for producing 20 Mw of electricity at a remote location is described. The reactor fuel consists of a light water uranyl sulfate solution which acts as its own moderator and coolant. The uranium is highly enriched (93% U/sup 235/). The primary considerstions for the design were simplicity and reliability of the components, automatic demand control and safe for any load change, full xenon override not required, possibility of construction within the immediate future, and economic operation not the cortrolling factor. Reasonably complete studies are presented for the reactor physics, safety, stability, chemistry, hent transfer, and operation of the system. (auth)
Date: August 1, 1957
Creator: Chapman, R.H.; Collins, H.L.; Dollard, W.J.; Fieno, D.; Hernandez- Fragoso, J.; Miller, J.W. et al.
Partner: UNT Libraries Government Documents Department

100-B Area flow analysis

Description: Results of experimental programs indicate that it might be desirable in the future to modify the existing reactors by replacing the aluminum process tubes with tubes made of a zirconium alloy. The zirconium tubes would be more corrosion resistant than the aluminum ones and would also be stronger at higher temperatures. These new tubes would have the same outer diameter as the present tubes (for ease of handling and in order to provide adequate graphite cooling) but would have a thinner wall (since zirconium alloy is both stronger and more expensive than aluminum). The inner diameter of the new tubes would, therefore, be greater than in the present tubes. In addition to the tube change, it might also be desirable to replace the existing solid fuel elements with those known as ``I&E`` alements. These pieces would be similar to the present elements except for a longitudinal hole which would allow the passage of cooling vater through the center. The element would then be Internally and Externally cooled, and would have a more uniform temperature distribution. The combination of the larger tube inside diameter and the central hole in the fuel element would result in reduced friction loss for the reactor cooling water with a resulting increase in flaw. The 100-B process water system was chosen arbitrarily for analysis and this analysis vas undertaken to determine if the process water system would be capable of providing the additional flow required by the modification.
Date: June 1, 1957
Creator: Bainard, W. D.
Partner: UNT Libraries Government Documents Department

183 B-C cross tie justification and scope

Description: After the present solid slugs are replaced with the new I&E elements` in the 105-B reactor, the friction loss for the reactor cooling water will be decreased with a resulting opportunity for increase in flow through the reactor. The amount that this flow could be increased is limited by the capacity of the B water plant as well as the reactor itself. It is possible that there will be a shortage of filtered water at 183-B during the critical periods of each year. To overcome this possible shortage of water it has been proposed to construct a thirty inch tie line from the 183-C reservoir to the 183-B clear-well to supply filtered water to 183-B by gravity flow from 183-C. This report presents justification and the scope of this project.
Date: December 2, 1957
Creator: Brinkman, L. B.
Partner: UNT Libraries Government Documents Department

ACCIDENT IN CONTINUOUS-DISSOLVER PILOT PLANT OF FLUORIDE VOLATILITY PROJECT ON MAY 15, 1957

Description: A series of explosions, estimated at five, occurred over a period of ten seconds within the continuous dissolver pilot plant, of the Fluoride Volatility Project on May 15, 1957. The explosive reactions occurred in the dissolver vessels as a result of violent chemical reactions between uranium and an interhalogen mixture. Just what the conditions were which triggered the explosions, have not been definitely established. Nevertheless, based upon the evidence which has been collected, several possible explanations, listed according to probability, are presented. A number of recommendations are included to be followed before operation of the pilot plant is resumed. These recommendations relate to additional laboratory research, equipment design, facility design, and use of a review committee. Safety rules for handling BrF/ sub 3/, BrF/sub 5/, ClF/sub 3/, and Br/sub 2/ are appended. (C.H.)
Date: July 10, 1957
Creator: Strickland, G.; Horn, F.L.; Johnson, R. & Dwyer, O.E.
Partner: UNT Libraries Government Documents Department

Action and Emission Spectra of the Luminescence of Green PlantMaterials

Description: The action and emission spectra of the delayed light emission from Chlorella, Nostoc, and spinach chloroplasts have been measured. The action spectra for Chlorella and for spinach chloroplasts are quite similar to the absorption spectra of these materials. The action spectrum for Nostoc, on the other hand, shows a relatively low activity for chlorophyll and carotenoids and a high activity for phycocyanin. The emission spectra of these materials demonstrates that the luminescence is the result of a transition between the first excited singlet state and the ground state of chlorophyll. Low-temperature studies suggest that the triplet state of chlorophyll is not involved at all in the luminescence of spinach chloroplasts. There is some indication that part of the light emitted from Nostoc is due to a phycocyanin transition.
Date: December 29, 1957
Creator: Tollin, G.; Fujimori, E. & Calvin, Melvin
Partner: UNT Libraries Government Documents Department

ACTIVATION CROSS SECTION OF Na$sup 23$ AROUND 3 KEV

Description: The activation cross section of sodium in the range from thermal energy through the 3-kev resonance is generally assumed to be given by the one-level BreitWigner formula, the GAMMA /sub gamma / being selected to describe correctly the known thermal absorption cross section. The contribution of this resonance to the resonance activation integral then turns out to be 0.12 barns. This somewhat indirectly inferred value is considerably larger than the value given by Dancoff et al., in an old paper, the latter value being based on experimental work. It is shown in the present memo, that Dancoff's actual measurements are quite consistent with the Breit-Wigner formula and the above mentioned GAMMA /sub gamma /. The discrepancy is a result of Dancoff's transition from the actual measurements to the resonance integral, this transition being based on data that is now obsolete. (auth)
Date: February 1, 1957
Creator: Ergen, W.K.
Partner: UNT Libraries Government Documents Department

Additional facilities to handle PUREX tank farm vapor wastes. Project CG-719

Description: The liquid high-level radioactive wastes from the separations plant are stored in large underground tanks where radioactive decay of the fission products in storage gives off heat. In the case of the 241-A underground storage tank farm, for Purex wastes, advantage is taken of this heat to self-concentrate the wastes. The present practice is to permit boiling and concentration in the storage tanks. The vapors given off from the boiling wastes are collected in a vapor header and passed through a deentrainment vessel and on to two contact condensers where the vapors are condensed and intermixed with waste cooling water. Samples taken of the waste vapors have shown a considerable amount of cesium{sup 137} present as well as other types of radioactive material carry over from the waste tanks. For this reason the contact condenser effluent is discharged to an underground crib 216-A-8. Underground disposal of the increasing volume of condenser effluent as larger waste volumes are accumulated in the underground tanks presents a critical problem which is further complicated by the desirability to transfer the condensate waste to new disposal facilities near the 200 West area. The intent of this report is to present the scope of the facilities required to reduce the volume of potentially radioactive condensate waste from the 241-A tank farm and to dispose of this waste through supplemental cribbing. An analysis of the 216-A-8 crib capabilities in relation to the projected flows clearly indicates that if other facilities to reduce the contaminated waste stream volume are not provided, an extensive and costly crib system will be required. The economical solution to the problem is to provide surface condensers to permit segregation of the condensed waste vapors from the cooling water, condensate collection and transfer facilities, and a new condensate disposal crib near the 200 West ...
Date: January 7, 1957
Creator: Wood, V. W.
Partner: UNT Libraries Government Documents Department

Additional Measurements on the Army Package Power Reactor Zero Power Experiments: ZPE-1 and ZPE-2

Description: During the course of the ZPE-2 experimental program additional measurements were performed under the Alco Products Research and Development program. Included in this program were the evaluation of various absorber section compositions and reactivity studies designed to facilitate analytical techniques. The results of these measurements are presented. (auth)
Date: November 15, 1957
Creator: Giesler, H. W.
Partner: UNT Libraries Government Documents Department

AIRBLAST OVERPRESSURE AND DYNAMIC PRESSURE OVER VARIOUS SURFACES

Description: Static overpressure and dynamic pressure versus time over surfaces processing different physical properties were measured on two tower shots, 6 and 12. On Shot 12, three surfaces were provided: the natural desert, a water surface consisting of a flooded area, and an asphalt surface. On Shot 6, desert and asphalt areas only were available. There were 123 channels of instrumentation installed for Shot 12, and 24 for Shot 6. From the data, a system of wave-form classification was devised for overpressure and dynamic-pressure- versus-time measurements. Incorporation of this system into data analysios indicates that it is possible for an ideal peak pressure to be identified with a nonideal wave form. Introducing both variables, wave form and peak pressure, into analyses reduces ambiguioties associoated with comparing results of different nuclear tests. The data show the effect of the nature of the surface upon airblast phenomena from a nuclear explosion. The effects of surface conditions upon shock phenomena are made more understandable by a review of temperature computatioons, using shock wave parameters in addition to an analysis based upon the arrioval time of the thermal pulse. A phenomenological discussion of precursor formation is presented, and comparisons are made using data from all known precursor-forming nuclear shots. Two Shot 12 dragforce measurements on the H Beams are presented and discussed. (auth)
Date: September 11, 1957
Creator: Sachs, D.C.; Swift, L.M. & Sauer, F.M.
Partner: UNT Libraries Government Documents Department

AN ALPHA COUNTER FOR UNIFORMITY MEASUREMENTS

Description: An ionization-type alpha counter is used to mensure the uniformity of thin coatings of alpha -emitting materials to plus or minus l% exclusive of statistical error. The (n, alpha ) reaction allows the counter to be used for measuring the uniformity of B/sup 10/ foils when a suitable neutron source is provided. (auth)
Date: July 1, 1957
Creator: White, F.A. & Sheffield, J.C.
Partner: UNT Libraries Government Documents Department

Alpha Ferric Oxide: Low Temperature Heat Capacity and Thermodynamic Functions

Description: ABS>The heat capacity of synthetic alpha ferric oxide was determined at 5 to 350 K. The experimental technique is described, and the heat capacity and molal thermodyamic functions are tabulated. The heat capacity vs. temperature is shown graphically. (J.R.D.)
Date: January 1, 1957
Creator: Gronvold, F. & Westrum, E. F., Jr.
Partner: UNT Libraries Government Documents Department

AN ANALYSIS OF POWER REACTOR FUEL REPROCESSING

Description: This report presents an analysis of the projected economies and processing capacity requirements for a power reactor fuel reprocessing industry based on the recovery of fertile and fissionable materials from presently proposed power reactors within tbe confines of the continental United 8tates for the next five to ten years. An analysis of the present general state of development of a technology required for such an Industry is given. A summary of results of power reactor reprocessing chemical and engineering development at Oak Ridge National Laboratory from July 1955 through December 1956 is given. (auth)
Date: March 27, 1957
Creator: Culler, F.L. Jr.; Blanco, R.E.; Goeller, H.E. & Watson, C.D.
Partner: UNT Libraries Government Documents Department

ANALYSIS OF PROMPT EXCURSIONS IN SIMPLE SYSTEMS

Description: A calculation, coded for a fast computer, is discussed that can fairly accurately match the experimental resuIts obtained from the LASL Lady Godiva prompt burst program. In particular, the code can predict the reactivity where the burst yield deviates from a linear dependence on reactivity and commences to rise much more rapidly. The burst width and maximum rates can be predictcd with reasonable accuracy. The calculation is applied to other simple systems, with the result that there appear to be general rules that one can use to relate the bursts of thcse systems. The calculation can predict what fraction of the energy of the system is converted into kinetic energy. It appears that step-function increases of reactivity of about 1O cents or less in homogeneous metal systems develop only a trivial explosive energy. The method was extended to hypothetical accidents in very idealized reactors. Two general models of reactor cores were considered. These are (1) a core of low density in which the characteristic o a threshold is important, and (2) a model in which the spherical shells are alternately normal density and void. The numbers quoted should be regarded only as establishing upper and lower limits. (auth)
Date: January 1, 1957
Creator: Stratton, W.R. & Colvin, T.H.
Partner: UNT Libraries Government Documents Department

Analysis of Stresses and Deflections in Top Support Grid, PWR Reactor. Final Report

Description: The top grid of the PWR reactor core assembly is treated as a simply supported circular plate. The theory of plate is applied to the grid umder mechanical loads. The thermal stress problem is analyzed by treating the plate as under combined action of a laterally distributed load and forces in the middle plane of the plate. The load distribution is calculated from the temperature variation over the grid. The thermal stress problem then is equivalent to two problems: one, of bendimg of plate; and, the other, a plane stress problem. The theoretical formulation for plates under nonuniform heating is developed by neglecting the effect of uneven expansion in the direction perpendicular to the plane of the plate. In replacing the partial dffferential equations by difference equations, the latter are modified to take into account the change in tbickmess and spacing of the grid webs near the boundary. Twentythree difference equations for the twenty-three stations in one octant of the grid are obtained for each second order partial differential equation. The difference equations are solved by assuming that the twisting moments and shearing stresses in the plane of the grid vanish at the boundary. The stresses and deflections due to mechanical loads and thermal expansion are then superposed. (auth)
Date: June 1, 1957
Creator: Yen, T. C. & Vining, R. E., Jr.
Partner: UNT Libraries Government Documents Department