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Heat Transfer to Superheated Steam

Description: Abstract: The physical property variation of superheated steam differs sufficiently from most other gases to warrant experimental investigation of heat transfer performance. Results are reported here of measurements made in a uniformly heated circular duct with steam at 1000 psi. The data agree very well with the expression use for design purposes, which is based on information in the literature for heating of other gases as well as steam. This work was a continuation of that performed under Task (Heat Transfer) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189, Project Agreement 13.
Date: May 1963
Creator: Sutherland, W. A. (William Alan), 1931-
Partner: UNT Libraries Government Documents Department

Nuclear Superheat Project. Internal Steam Separation Development of Radial Vane Steam Separators

Description: This technical report describes the development, design, operation, and performance of a full-circle, radial-vane steam separator for the boiling water section of a nuclear superheat reactor. Steam-water tests of this model have demonstrated that is has vane capacity in excess of that required for the 300-Mx(e) separate superheat reactor and for the 300-Mw mixed spectrum superheat reactor. It is proposed that the vane capacity requirement of the 600 Mw(e) separate superheat reactor may be attained by increasing the nozzle length.
Date: May 31, 1963
Creator: Moen, R. H.
Partner: UNT Libraries Government Documents Department

Two-Phase Pressure Losses Quarterly Progress Report: Fifth Quarter, February 12, 1963 - May 12, 1963

Description: Technical report describing that void measurements were made in the 1/2-inch by 1-3/4-inch rectangular channel, for both flow up and flow down, at pressures of 600, 1000, and 1400 psia, and at various flows and quantities. Results at 1000 psia and 20 percent quality show that for the lowest flow both the void distribution and the average void are much different for flow down than for flow up, the void fraction for flow down being much higher. However, when the flow is increased both the void distribution and average void for flow down tend to approach the corresponding values for flow up. At 1000 psia, both flow up and flow down, the void fraction for 5 percent quality increases gradually from the wall to the center of the channel, and peaks at the center. At 20 percent quality, the void fraction increases abruptly from the wall and tends to be constant over the middle 65 percent of the channel. the void fraction for flow down is always greater than for flow up, other things being equal.
Date: June 1, 1963
Creator: Janssen, E. (Engineer). & Kervinen, J. A.
Partner: UNT Libraries Government Documents Department

Applications of Strain Cycling Considerations to Superheat Fuel Design

Description: A potential performance limitation of superheat fuel is the susceptibility of the fuel cladding to low cycle fatigue failure. Two simplified analytical methods are presented to estimate the cyclic lifetime of circular superheat fuel cladding. One failure relation is based on a displacement method. The other failure relation is based on a stress method. These relations were compared with data from the literature, and with data involving damage obtained by Reynolds. A recommended design procedure involving the relations is presented. The technique was applied to the SADE 4B experiment with moderate success. These cycling relations involve only mechanical damage imposed by cycling, with a modification for additional damage caused by radiation; they do not include any other potential performance limiting mechanisms, such as stress corrosion, which are normally factored into the over-all fuel design. This work work done under Task C (Materials Development) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189 - Project Agreement 13.
Date: June 1963
Creator: Rieger, G. F.
Partner: UNT Libraries Government Documents Department

A Program of Two-Phase Flow Investigation Quarterly Report: First Quarterly Report, March-June, 1963

Description: Task A: Modification and Preparation of Experimental Facility. Facility engineering and layout is about seventy-five percent complete. Task B: Design and Construction of Test Sections. The major dimensions and characteristics of the metal and glass test sections have been calculated. One feasibility test of the electrically conducting coating on samples of glass tubing has been completed. Task C: Design and Construction of Test Stand, Task E: Pressure and Temperature Instrumentation for Test Section and Task F: Power Supply for Test Section. Preliminary engineering has been initiated on these tasks. The planned approach has been defined in each case. For Task E the transducer specifications have been defined and quotations on and/or sample units of the transducers have been requested. Tasks C and F can proceed with detailing as soon as drafting on Task B is about 50 percent complete. This point is scheduled to be reached during the first part of July. Task D: Void Fraction Instrumentation. The requirements for the x-ray instrumentation have been considered in the course of Task B and the x-ray power supply is presently on hand. The detailed engineering effort on this task is not scheduled to begin before July.
Date: June 24, 1963
Creator: Staub, F. W. & Zuber, N.
Partner: UNT Libraries Government Documents Department

Design and Fabrication of Coextruded Stainless Steel Clad UO2 Fuel Rods

Description: A process was developed in which stainless steel-clad UO2 fuel rods are fabricated by high-temperature coextrusion. The process has a potential of being a more economical method for the preparation of stainless steel-clad UO2 fuel rods than the conventional pellet process. Consequently, it was considered advantageous to evaluate the irradiation characteristics of fuel rods fabricated in this manner. Therefore, 24 coextruded fuel rods were manufactured for evaluation in a reactor. The required amounts of UO2 and clad were soaked in separate containers at 1875 and 760 degree C, respectively. The containers were removed from their respective furnaces and were coextruded in one pass. A force of 450 to 475 tons was used, and a reduction ratio of 18 to 1 was obtained. The coextruded rods were cut to the approximate length, and the ends were sealed with an acid-resistant tape. The carbon steel can covering the stainless steel clad was removed by immersion in 1:1 nitric acid for 20 minutes. The rods were visually inspected, the specified lengths of clad and fuel were obtained by machining, and the correct diameter was obtained by belt sanding. The fabrication of the fuel rods was completed by inserting the plenum support tubes and welding in the end plugs. Nineteen of these fuel rods were sent to the Atomic Power Equipment Department (APED) for irradiation in the Vallecitos Boiling Water Reactor (VBWR). The irradiation of 12 of these rods was begun in August 1961, while irradiation of 3 rods was begun in July 1962. The irradiations will continue until an average burnup of at least 10,000 Mwd/t is achieved by some of the fuel rods.
Date: June 1963
Creator: Baroch, C. J.
Partner: UNT Libraries Government Documents Department

Maritime Loop Irradiation Program, S-I-5-B-M Fuel Irradiation Water Chemistry, Final Report

Description: Introduction: The purpose of this technical report is to review the water chemistry methods and equipment developed for use with the Maritime Loop Irradiation Program conducted in the General Electric Test Reactor (GETR) from December 2, 1960 to July 19, 1962. Special emphasis is given to areas having general application to other high purity water systems. The Appendix includes a discussion of specific conductivity and pH in high purity water systems. A major section of this report is devoted to a review of gross activity levels on coupons of two different surface finishes exposed in the loop coolant system for various time intervals. A major objective of the chemistry program was to select or develop analytical methods such that the analyses could be performed at the loop location by technical personnel who normally operate the loop. By this means, frequent samples were obtained and analyzed directly thus providing close monitoring and control of the loop water chemistry at minimum expense.
Date: August 1, 1963
Creator: Danielson, D. W.; Gilbert, R. S. & Panter, G. E.
Partner: UNT Libraries Government Documents Department

Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Fifth Quarterly Progress Report, April 1-June 30, 1963

Description: Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. The development of pulsed neutron source techniques for large power reactors has led to a new theoretical model recently developed by E. Garelis and J.L. Russell, Jr. The theory is presently based on a bare, one-group model with m-delayed precursors and takes all spatial modes into account. Results indicate, however, that the application of this model is much broader. Experiments were designed and carried out to both verify this new theory and to demonstrate the performance of the experimental hardware in a large power reactor.
Date: July 15, 1963
Creator: Garelis, Edward. & Meyer, P.
Partner: UNT Libraries Government Documents Department

In-Core Instrumentation Development Program, Detectors for In-Core Power Monitoring

Description: Introduction: The object of Project Agreement 22, Task 1, is to develop improved detectors which can operate up to 1000 F for in-core power monitoring. Several ideas have been developed to achieve this goal: (1) root mean square fluctuation voltage measurement of ion chamber signals, (2) thermocouple-type detectors, and (3) fabrication developments.
Date: June 1963
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department

Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: April 1 - June 30, 1963

Description: A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. All program efforts have been temporarily deferred except for those associated with the irradiation of the program fuel element in the VBWR. The program fuel element was exposed to a burnup of 831 MWD/T during the quarter which brings the total to 3165 MWD/T. Applying the same scale factor between logged exposure and Ce-Cs analysis of the first fuel sample gives a corrected exposure of 3774 MWD/T.
Date: July 15, 1963
Creator: Robkin, M. A.
Partner: UNT Libraries Government Documents Department

Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 3

Description: The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Date: July 1, 1963
Creator: Sorlie, T.
Partner: UNT Libraries Government Documents Department

Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Seventh Quarterly Report, April-June 1963

Description: Quarterly report discussing progress on the Fast Ceramic Reactor Development Program, "an integrated analytical and experimental program directed toward the development of fast reactors employing ceramic fuels, with particular attention to mixed plutonium-uranium oxide" (p. 1).
Date: July 1963
Creator: Leitz, F. J.
Partner: UNT Libraries Government Documents Department

In-Core Instrumentation Development Program, Telemetering Transmitters for In-Core Power Monitoring Final Report

Description: Abstract: This technical report covers the development work conducted during a planned program with the U.s. Atomic Energy Commission, Contract AT(04-3-189, Project Agreement 22, directed toward the development of high temperature, nuclear radiation resistant, telemetering devices. The development program is devoted to: (1) investigation and selection of two possible telemetering devices, and electromechanical commutating switch and an AM oscillator employing TIMM circuit elements, (2) procuring the electromechanical commutating switch to specification, (3) building and operating a TIMM oscillator, and (4) temperature testing of both devices. A resistance-coupled Wien-bridge sine wave TIMM oscillator was build and tested both as an oscillator, and in combination with other oscillators to simulate a telemetering system. An electromechanical commutating switch rated for 350 F operation, instead of 700 F as originally specified, was procured and tested. The drive motor and gear reduction unit which is designed to drive the commutating switch, is rated for 750 F operation and designed to operate in an nuclear reactor radiation environment of 1 x 10(17) nvt and 1 x 10(10) R.
Date: July 1963
Creator: McQueen, A. H.
Partner: UNT Libraries Government Documents Department

Specific Zirconium Alloy Design Program Quarterly Progress Report: Fifth Quarter, April - June, 1963

Description: A program is in progress for the design of a zirconium base alloy for steam service as nuclear fuel cladding. Thirty-one alloys selected for study of corrosion rate, hydriding rate and hydrogen embrittlement are in test. The corrosion testing of 1800 coupons to 3000 hours at at 300, 400, and 500 degrees C in refreshed steam has been completed. Statistical data analysis of the corrosion results are reported and alloys showing better corrosion performance at all test temperatures than that for Zircaloy-1 are discussed. Preliminary data for hydrogen uptake after long exposures at 400 and 500 degrees C are presented; the uptake for alloys showing the best corrosion performance is discussed. Post-corrosion mechanical property measurements are also reported along with the preliminary results of x-ray diffraction and metallographic studies relating to hydrogen embrittlement. A wide variation in resistance to embrittlement at a given hydrogen level was observed and can be tentatively correlated with original ductility, crystallographic texture, and hydride platelet orientation. The testing of a second round of ten alloys is also in progress. Studies concerning the mechanism of corrosion and hydriding in zirconium alloy are also reported. The results of recent neutron activation analyses of stripped corrosion films are presented. Oxygen diffusion through doped non-stoichiometric ZrO2 is now proceeding following earlier difficulties in sample preparation. Work on hydrogen overvoltage and electrochemical potential of inter-metallic phases was previously completed and reported.
Date: July 1, 1963
Creator: Klepfer, H. H.; Jaech, John L.; Blood, R. E. & Douglass, D. L. (David Leslie), 1931-
Partner: UNT Libraries Government Documents Department

Environmental Testing of a B4C-Nl Prototype Control Rod

Description: Summary: A prototype control rod containing absorber plates made from an electro- deposited dispersion of boron carbide in nickel was tested in the VBWR. It was exposed to the reactor environment of 545 degree F boiling water and thermal neutron fluxes (perturbed) which ranged from 0.6 to 1.1 x 10/sup 13/ nv for 2236 hours over a period of six months. The maximum B/sup 10/ burnup achieved during the test period was 1.8 percent. After irradiation, the rod was examined. The results of the examination are summarized below: (1) The B/sub 4/C-- Ni plate assembly did not undergo significant dimensional changes during irradiation. (2) Numerous blisters developed on both the outer and inner surfaces of three of the four plates. Blistering was more severe on the outer surface than on the inner, and was most severe in a large region located in the lower half of plate 4. Metallographic examination revealed that the blisters were located only in the 2- mil protective nickel overlay covering the B/sub 4/C-- Ni dispersion. It was concluded that they formed from the buildup of gas pressure at the Ni: Ni-- B/sub 4/C interfaces, rather than from corrosion attack. Helium from the B/sup 10/(n alpha )Li/sup 7/ reaction probably contributed to this pressure. However it is conjectured that the major gas was very likely hydrogen, possibly generated and dissolved in the nickel during electroplating and then released to defects at the Ni: Ni--B/sub 4/C interface during reactor exposure. The variation in the degree of blistering among the four plates of the prototype indicated that the blistering was related to variations in the fabrication process. Failure of the nickel overlay was not observed in any of the blisters examined metallographically, and the underlying B/sub 4/C-- Ni appeared to be in good condition. (3) Evidence of corrosion ...
Date: October 15, 1963
Creator: Megerth, F. H. & Zimmerman, D. L.
Partner: UNT Libraries Government Documents Department

Transition Boiling Heat Transfer Program; Second Quarterly Progress Report, April - June 1963

Description: Introduction: The Transition Boiling Heat Transfer Program is sponsored jointly by the USAEC and Euroatom and is being conducted by the General Electric Company. The work commenced on this program February 11, 1963. The objective of this program is to perform basic investigation and measurement of the transition boiling regime in high pressure bulk boiling water flows, with particular emphasis i the high range of steam qualities.
Date: July 1, 1963
Creator: Quinn, E. P.
Partner: UNT Libraries Government Documents Department

Sodium Mass Transfer. [Part] 8. Corrosion of Stainless Steel in Isothermal Regions of a Flowing Sodium System

Description: Technical report describing an analytical investigation made on the mechanism of the "downstream" effect in the corrosion of stainless steel in sodium. A mechanism of iron alloy corrosion is assumed in which the controlling rate is diffusion of iron-oxygen species, probably a FeO-Na2O complex. A mathematical model is developed for the corrosion, and the predicted results agree with the experimental data. The corroding species is probably present in sodium at concentrations of ~10(-8) g Fe/g Na.
Date: February 1964
Creator: Mottley, J. D. & Epstein, Leo F.
Partner: UNT Libraries Government Documents Department

General and Localized Corrosion Studies of Type 300 Series Austenitic Stainless Steels in Simulated Superheat Reactor Environment

Description: The following conclusions are based on the out-of-pile general corrosion and localized attack studies completed to-date on several 300 series stainless steels: (1) Utilizing a sodium chloride-cycle test that produces a type failure that can occur in a superheat reactor system, Types 347 and vacuum-melted 304 SS have failed while vacuum-melted 310 SS was acceptable. (2) An improved chloride cycle test utilizing ferric chloride as the additive has been developed that produces an intergranular type failure similar to that experienced in the fuel cladding failures in the SADE and ESADE facilities. types 304 and 315 SS have failed in the test. (3) Present methods of ultrasonic testing will find through cracks but are not completely dependable for assessing lesser degrees of intergranular attack. (4) It is hypothesized that a definite interplay exists between chemical attack and stress. The application of stress will orient intergranular attack preferentially in a direction perpendicular to the stress.
Date: July 1963
Creator: Pearl, W. L.; Gaul, G. G. & Wozadlo, G. P.
Partner: UNT Libraries Government Documents Department

Physics Design of the Mixed Spectrum Critical Assembly

Description: Summary: The Mixed Spectrum Superheater (MSSR) is an integral superheater reactor in which boiling occurs in an annular Boiling Water Reactor section and steam in superheated in an unmoderated fast section in the center. A Mixed Spectrum Critical Assembly (MSCA) to be operated at the Vallecitos Atomic Laboratory has been designed to mock up a 75-150 MWe prototype MSSR. The principal experimental measurements aimed at proving the feasibility of the MSSR concept include power distribution, Doppler effect, flooding effects, distribution of reactivity, control rod worths, and the effect of the control system on the power distribution.
Date: August 1963
Creator: Reynolds, A. B.
Partner: UNT Libraries Government Documents Department

The Measurement of Fission Gas Pressure in Operating Fuel Elements: Post-Irradiation Examination

Description: Summary: Two UO2-filled stainless steel clad fuel rods in which fission gas pressure was measured during irradiation have been subjected to post irradiation examination. Results of free gas analysis and metallographic examination are in general agreement with observed pressures reported previously. Calculated fuel surface temperatures based on extent of fuel recrystallization indicate that in a one-half inch diameter fuel rod with 0.014 inch diametral clearance operated at a maximum heat flux of 531,000 Btu/hr-ft, gap conductance increased with increasing heat flux. An analysis of void configuration indicates that pressure is more sensitive to as-fabricated void volume and changes in this volume resulting from fuel expansion than to fuel central temperature. The decreases in effective void volume with increasing fuel temperatures becomes more significant as initial void volume decreases, and excessive fission gas pressures may be developed in fuel rods operated at high fuel temperatures unless adequate expansion volume is provided in fabrication.
Date: September 20, 1963
Creator: Reynolds, M. B.
Partner: UNT Libraries Government Documents Department