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SNAP Aerospace Safety Program Quarterly Technical Progress Report, July- September 1962

Description: Statements are presented concerning project objectives, major accomplishments in FY 1962, progress during the report period, evaluation of effort to date, and next report period activities. The subject material covers reactor separation and fuel element ejection, reactor transient and excursion tests, reactor end-of-life shutdown devices, fission product release studies, critical configuration tests, mechanical and thermochemical effects, and fuel element burnup and fission products dispersal. Major emphasis for the period was concentrated on the design of the SNAPTRAN 2/10A-1 and -2 machines and the models for the Reentry Flight Demonstration (RFD-1). Planning of the SNAPTRAN experimental program was undertaken. Analytical effort was concentrated on developing the several computer codes required for interpreting and understanding the data acquired or to be acquired from the experiments. The majority of the Phase I mechanical and thermochemical effects test program was completed. A few fission product releases were conducted at the NRTS. Necessary test apparatus for the water immersion critical experiment was designed and fabricated. (auth)
Date: March 20, 1963
Partner: UNT Libraries Government Documents Department

Reactor Kinetics: A Bibliography

Description: ABS>A bibliography covering the material listed in Nuclear Science Abstracts and Abstracts of Classified Reports up to July 1958 is presented. It is divided into 39 sections with author and report number indexes and a glossary of abbreviations. Each subject group has the entries by author onder, or by title (if no author listed). Each entry has a number so as to assist the location of reports from either the author or number indexes. While this biblicgraphy includes classified reports (indicated by AEC Classified), there are no classified titles. Publication dates of reports are given, when available, except where a date shows in the title, such as in progress reports. The report number index may be of special value to those researchers who desire all of one company's reports on this subject. (auth)
Date: November 1, 1959
Creator: Bloomfield, M. & Bennet, F.G.
Partner: UNT Libraries Government Documents Department

Reactor Kinetics Quarterly Progress Report for January-March 1957

Description: Investigation of the energy coefficient of radiolytic gas production for the KEWB I reactor as a function of core temperature shows the coefficient to be independent of temperaturc above about 30 C. The temperature coefficient of reactivity is found to increase markedly with temperature and reach a value of 0.031%/ C at 90 C with a core pressurc of 71 cm Hg. A void coefficient of reactivity was determined in the certral exposure tube of the reactors and is related to the void coefficient derived from the temperature coefficient measurements. Investigations of the transient power oscillations show a decrease in the period of oscillation and an increase in the damping of the oscillation with increasing reactivity input. An evaluation of the ''boron-bubble'' experiment and its extension to more complex geometries was made in conaection with the determination of ''absolute reactivity.'' This study indicated a possible solution to the difficulties involved in the experimental determination of the absolute reactivity as contrasted with the reactivity in dollars observable in kinetic experiments. A method was devised for determining the transfer function of a reactor from power fluctuations during steady operation by treating the fluctuations as the response of the system to an inherent random or stochastic driving force. A power series solution of the one-group reactor kinetics equations was obtained for the rcactivity as a prescribed cubic function of time. (For preceding period see NAASR-2134.) (auth)
Date: January 1, 1959
Creator: Remley, M.E. ed.
Partner: UNT Libraries Government Documents Department

Reactor Safety Quarterly Progress Report for January-March 1957

Description: ABS>Seven special safety elements of the Mark II design were rebuilt and are ready for resumption of lifetime testing. Precision-cast parts were received for ten Mark IV highpressure chambers; one set was assembled and pressdre tested to 2000 psi. Two special assemblies of the NAA 1093 experiment were completed and are being shipped to Hanford. Conductivity measurements on stainles s steel were continued in an effort to determine the cause of erratic behavior of the variable-cooled trigger. A vacuum chamber was used in am effort to eliminate local heat loss due to convection. A study of corrosion rates of various container materials in liquid-metal poisons was umdertaken. Experiments with the differential-pressure device on the rate of pressure rise thnt can be achieved by electrically heating a closed gas volume were continued. An exponential generator is being constructed to enable simulation of reactor periods from 30 msec to one sec. Three types of research reactor safety device were chosen for further study: the pyrophoric, the double-diaphragm, and the electronic-explosive. Preparations are in process for demonstrating models of the latter two devices in the KEWB facility. (For preceding period see NAA-SR1954.) (M.H.R.)
Date: December 1, 1957
Creator: Miller, N.C. ed.
Partner: UNT Libraries Government Documents Department

Recent Developments in the Technology of Sodium-Graphite Reactor Materials

Description: Experimental results on core structural materials, moderator, and fuel materials are discussed. Data are given on grain growth of Zr and thermal conductivity, thermal expansion, and temperatures of various types of graphite. (M.H.R.)
Date: October 31, 1958
Creator: Carter, R.L. & Eichelberger, R.L.f Siegel, S.
Partner: UNT Libraries Government Documents Department

The Response of a Water Boiler Reactor to Very Fast Power Transients and Linearly Increasing Reactivity Inputs. Water Boiler Excursions With an Initially Filled Core

Description: A report is made on the Kinetic Experiment on Water Beller Program. The purpose of this program is to examine the dynamic behavior of homogeneous research reactors to obtain the information necessary for the evaluation of the nuclear safety of such reactors. Step inputs of reactivity were systematically increased and the first test core, a spherical core designed for stable power operation at 50 kw, was examined under conditions of 4% reactivity release This is the maximum normally installed in such reactors A 4% reactivity release places the reactor on a 2 millisecond stable period and leads to a peak power of 530 Mw. This represents the fastest intentional power excursion of any thermal reactor. The reactivity released is more than twice that which any other has withstood without damage. The maximum pressure in the system for this transient was a sharp pressure peak of 370 psia. This pressure is well below that required to cause yield of a typical water boller core. (A.C.)
Date: September 1, 1958
Creator: Stitt, R.K.; Gardner, E.L.; Roecker, J.H.; Wimmer, R.E. & Hetrick, D.L.
Partner: UNT Libraries Government Documents Department

Safety Device Tests in KEWB I

Description: The feasibility of the electronic-explosive safety device was demonstrated in KEWB-I reactor excursions. It was shown to be a fast, efficient safety system for reactor protection. The safety devices inserted 1.5% negative reactivity into the core region of the KEWB-1 reactor to produce rapid shutdown. The system response time, i.e., that time interval between trip and insertion times, was found to be 4.2 msec. The complete freedom in choice of trip level and the flexibility of device geometry makes the system applicable to many types of reactor and critical assembly. The system does not offer the ultimate in reactor safety because it is not completely self-contained, but it is a practical, workable system for reactor protection. (auth)
Date: January 28, 1958
Creator: Weeks, C. C. & Fitch, S. H.
Partner: UNT Libraries Government Documents Department

A Siphon Break as a Blocking Valve

Description: An experiment was conducted io determine the feasibility of using the breaking of a siphon as a quick-acting means for stopping sodium fiow following a loss of pump power. A 2-in. pipe system with a high-speed free-surface centrifugal pump was used in this investigation. Runs were made with sodium at 500 and 940 deg F, cover gas at various pressures up to 10 psig, and Reynolds numbers up to approximately 360,000. The siphon-break was established as an effective method for rapid flow stoppage; however, a brief reversal of flow follows the initial flow stoppage. An expression for the flow transient following the breaking of the siphon was derived which agreed reasonably well with experimental results. (auth)
Date: October 15, 1959
Creator: McDonald, J. & Marten, W.
Partner: UNT Libraries Government Documents Department

Sodium Graphite Reactor Materials Survey

Description: >The materials problems associated with the present sodium graphite reactor system have generally been approached by using existing knowledge and data to meet the proposed operating conditions. This discussion reviews the general reactor concept and the specific materials used for the major reactor components: (1) shielding materials; (2) core materials; and (3) sodium cooling system materials. In each case, the materials problems and the materials used to minimize or eliminate these problems are described. Economical nuclear power is currently dependent on the flow of improved materials for high temperature use in high radiation fields. Rapid progress is being made in this respect. (auth)
Date: September 15, 1959
Creator: Hayward, B. R.
Partner: UNT Libraries Government Documents Department

The Sodium Graphite Reactor Power Plant for CPPD

Description: The plant arrangement, component design, and the functions of various systems are described and illustrated. Relative estimated costs of the systems and major components are indicated. The reactor core is designed around requiremouts for 254 thermal megawatts, 950 deg F maximum sodium temperature, stainless steel clad graphite moderator blocks, and low enrichment (0.015 to 0.04 U/sup 235/) uranium fuel elements. The fuel cycle is described for the possible fuel elements. The fuel cost factors are discussed. Burn-up limitations encountered for metallic fuel in the SGR temperature range indicate UO/sub 2/ the more desirable choice. The estimated cost of electrical energy associated with the UO/sub 2/ fuel is given. (auth)
Date: October 31, 1958
Creator: Olson, R.L.; Gerber, R.C.; Gordon, R.B.; Ross-Clunis, H.A. & Stolz, J.F.
Partner: UNT Libraries Government Documents Department


Description: An experiment was conducted in the SRE to measure temperatures and neutron flux levels in and near a boron-containing simulated control rod. The data are being used to check analytical methods developed for prediction of control rod heat generation rates and maximum temperatures in this type of control rod in the Hallam Nuclear Power Facility. The maximum observed temperatures with a reactor power level of 20 Mw were 1363 deg F for a boron-- nickel alloy ring having a 0.105-in. radial clearance with the thimble and 1100 deg F for a boron -nickel alloy ring having a 0.020-in. radial clearance. The maximum temperature difference between the coolant and the control rod was 473 deg F. It is concluded that the expected greater heat generation rates in the Hallam reactor would prohibit the use of boron-containing absorber materials in a combined a him-safety rod. (auth)
Date: June 1, 1959
Creator: Arneson, S.O.
Partner: UNT Libraries Government Documents Department

Fuel Programming for Sodium Graphite Reactors

Description: The effect of fuel programming, i.e., the scheme used for changing fuel in a core, on the reactivity and specific power of a sodium graphite reactor is discussed Fuel programs considered Include replacing fuel a core-load at a time or a radial zone at a time, replacing fuel to manutain the same average exposure of fuel elements throughout the core, and replacing and transferring fuel elements to maintain more highly exposed fuel in the center or at the periphery of the core. Flux and criticality calculations show the degree of power flattening and the concurrent decrease in effective multiplication which results from maintaining more exposed fuel toward the core center. Corverse effects are shown for the case of maintaining more exposed fuel near the core periphery. The excess reactivity which must be controlled in the various programs is considered. Illustrative schedules for implementing each of these programs in an SGR are presented. (auth)
Date: October 15, 1959
Creator: Connolly, T.J.
Partner: UNT Libraries Government Documents Department

High-Strength Zirconium Alloys

Description: The properties of zirconium alloyed with aluminum tin, and molybdenum were investigated. Using reactorgrade zirconium sponge, 11 zirconium-base alloys were double arc-melted and cast into 6-in.-diam. ingots weighing 35 lb each. By such standard hot working procedures as extruding and rolling, the ingots were converted to 1/8-in.-thick strips. The extruded and rolled products were used for a variety of evaluation studies which included corrosion thermal conductivity, tensile, and creep tests. The alloys demonstrated short-time elevated temperature strength properties equal to or greater than type-304 stainless steel. Their corrosion resistance in sodium, at 1000 deg F, compares favorable with that of unalloyed zirconium. The creep resistance and the thermal conductivity were found to be less than those for type-304 stainless steel, but adequate for nuclear reactor application. (auth)
Date: July 15, 1959
Creator: Wagner, R.K. & Kline, H.E.
Partner: UNT Libraries Government Documents Department

HNPF Cold Trap Evaluation

Description: Two designs of sodium cold traps for the HNPF have been subjected to full scale tests, Performance features that were investigated include oxide removal efficiency, oxide capacity, pressure drop characteristics, economizer effectiveness, and temperature profiles, Results indicate that both designs should perform satisfactorily in the Hallam plant, (auth)
Date: December 15, 1959
Creator: Cygan, R.
Partner: UNT Libraries Government Documents Department

Preliminary Design Study for a Sodium-Graphite-Reactor Irradiation Facility

Description: The results of an investigation to integrate a Na/sup 24/ irradiation processing facility with an operating sodium graphite reactor are presented. An irradiation facility incorporated into a reference SGR (Hallam Nuclear Power Facility, Hallam, Nebraska) is described. Development of the facility application, preliminary design criteria and capital and operating costs are discussed. Recommendations for further development of the technology and economics of this type of irradiation facility are included. (auth)
Date: January 31, 1959
Creator: Thompson, D.S. & Benaroya, V.
Partner: UNT Libraries Government Documents Department

Bubble Formation: A Bibliography

Description: Bubble phenomena have been given a new meaning with their study in relation to the kinetic behavior of reactors. Prior to their study in relation to physics, the bulk of work on bubble phenomena concerned naval engineering problems of behavior in cavitation and water entry behavior. This bibliography is intended to fill the need of the reactor physicist as well as the naval engineer. An attempt has been made to include all available references on bubble phenomena and associated effects. A subject index has been purposely omitted. It is felt that the breakdown in content headings is sufficient to ascertain areas of interest. There will be overlapping of headings and to find all possible entries, a search through the headings may be desirable. To increase the usefulness of this bibliography the location of an abstract has been cited wherever possible following the reference. Classified reports are included; however, their titles contain no classified information. Sources used in compiling this bibliography are: Chemical Abstracts, Industrial Arts Index, Applied Mechanics Review, Nuclear Science Abstracts, the AEC Abstracts of Classified Literature, the AEC card catalogs available at Atomics International, and the bibliographic services of Armed Services Technical Information agency. (auth)
Date: June 30, 1958
Creator: Bloomfield, M.; McElroy, W.N. & Skinner, R.E.
Partner: UNT Libraries Government Documents Department

Calibration of Omre Fuel-Element Surface Thermocouple Assembly

Description: Studies were made to determine the actual surface temperature of OMRE fuel elements if the thermocouple were not present. Chromel-alumel thermocouples are being attached to the fuel plate cladding of Type 304 stainless steel. These wires are in contact with the coolant stream. Heat transfer from the thermocouple junction, by conduction along the lead-wires and by forced convection to the coolant, produces a lowering of the surface temperature in the region of the junction which results in an error in surface temperature measurement. (W.L.H.)
Date: March 12, 1959
Creator: Sudar, S.
Partner: UNT Libraries Government Documents Department

Casting Development for Uranium-Molybdenum Alloy Shapes

Description: The casting of shapes of uranium--molybdenum metal of varying sizes and thicknesses from a molten charge has been successfally accomplished with specificially designed graphite distributors and molds. Solid cylinders, hollow cylinders, and flat plate shapes were cast in gang molds. As many as 35 solid cylinders have been cast simultaneously. All castings had smooth surfaces, and solid shapes were cast to 0.006-in. tolerance on all dimensions except length. (auth)
Date: November 15, 1959
Creator: Binstock, M. H. & Stanley, J. A.
Partner: UNT Libraries Government Documents Department

Danger Coefficient Measurements Using a Water Boiler Reactor

Description: The use of a water boiler reactor as a danger coefficient test instrument has been investigated, with emphasis on the testing of reactor materials for neutron absorbing impurities and the testing of uranium for enrichment variations. After calibration of the WBNS course control rod, tests were made on a variety of materials including beryllium, boron trifluoride, iron and steel, natural uranium metal and oxide, and uranium metal of various enrichments. Several experiments were conducted to evaluate the sources and magnitudes of errors in danger coefficient testing. On the basis of the data obtained from sample testing and auxiliary experiments, the sensitivity of the WBNS for absorption and enrichment testing was determined, as well as other constants relating to danger coefficient measurements. The Water Boiler Neutron Source has been found to be well suited to danger coefficient testing. Small absorption differences between samples of similar geometry and nuclear propenties can be determined to a standard deviation of 0.002 cm/sup 2/. The danger coefflcient technique using the WBNS compares favorably with other methods of absorption detecting methods, such as the shotgun test or chemical analysis. Enrichment differences between natural uranium samples of only 500 grams can be detected to plus or minus .0008 per cent U/sup 235/ by weight, and between 3 per cent enriched samples to plus or minus 0.0028 per cent U/sup 235/. In enrichment testing relative values can be found to an accuracy equivalent to those obtained from spectrographic analyses. With improvements in the WBNS core arrangement so that larger samples could be handled, and relocation of the glory hole to the flux center of the core, greater sensitivity could be obtained. The WBNS compares favorably to other reactors used for danger coefficient work (auth)
Date: March 1, 1956
Creator: Engholm, B.A.
Partner: UNT Libraries Government Documents Department