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Yankee Atomic Electric Company Research and Development Program: Quarterly Progress Report: April 1, 1958 to June 30, 1958

Description: Quarterly report describing the technical research and development work accomplished by the Yankee Atomic Electric Company Research and Development Program, and documenting progress made on the group's various projects and goals.
Date: July 30, 1958
Creator: Coen, L. H. & Garbe, R. W.
Partner: UNT Libraries Government Documents Department

Yankee Atomic Electric Company Research and Development Program: Quarterly Progress Report: October 1, 1958 to December 31, 1958

Description: Quarterly report describing the technical research and development work accomplished by the Yankee Atomic Electric Company Research and Development Program, and documenting progress made on the group's various projects and goals.
Date: February 16, 1959
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department

Design Study of Portable Thermoelectric Nuclear Systems

Description: Design studies were performed and costs were estimated for an air transportable, 10 Mw(t), pressurized light water thermal circulation reactor, combined with a direct conversion thermoelectric generator and static electrical inversion equipment. This TCR-TE'' concept appears to have potential for ultimate use as a remote unmanned power station. Based on an extrapolation of present reactor technology and on assumed thermoelectric materials properties forecasted to January 1, 1963, a net a-c electrical output of 315 Kw is estimated, assuming the use of 80 deg F local water for cooling purposes. An alternate concept using 80 deg F air for cooling produces 271 Kw, net. These electrical output figures can be improved significantly through a recommended research and development effort. The design and construction of a prototype plant is also recommended. (auth)
Date: July 1, 1961
Creator: Chajson, L.; DelCampo, A. R. & Kellogg, H.B. et al
Partner: UNT Libraries Government Documents Department

Development of Equations for Analog Computer Studies to Size the Reactor Plant Pressurizer

Description: The assumptions and equations used to conduct reactor plant load trnnsient studies on the analog computer are presented. The study was performed to determine the magnitude of reactor cooling water temperature and volume variations caused by secondary plant load transients, and to establish the size of the pressurizer which would be capable of limiting the cooling water pressure variations caused by the volume surges. (auth)
Date: October 15, 1958
Creator: Lyman, W. G.
Partner: UNT Libraries Government Documents Department

The Effect of Local Boiling on Pressure Drop and Flow Distribution in the Yankee Reactor Core

Description: Where local boiling can occur in a semi-open heterogeneous reactor core consisting of parallel flow channels, differences in the terms comprising the total pressure drop in the local boiling regions and in parallel adjacent regions canse flow redistribution in the reactor. This report covers the study of the effect of local boiling on the friction factor and momentum components of the total pressure drop. The local-boiling to single-phase friction factor ratio has a maximum value of 1.42 when a linear relationship is assumed between the enthalpy of the coolant and the friction factor ratio. Analytical and gaphical methods for determining the region of the core in local boiling are developed. For a cylindrical reactor core similar to the Yanke reactor core the maximum possible region of the core in which local boiling should occur would be an ellipsoid whose surface is the locus of points of local boiling. In the Yankee core the maximum possible volume in local boiling is 10.4 cubic feet and 14.4 cubic feet in the first and second parameter studies, respectively. These volumes are 4.5% and 5.5% of the total volume of the core. The nature and mechanism of local boiling is discussed and its relation to the single phase and true two-phase pressure drop is reviewed. Flow redistribution out of the hot channel due to local boiling is 17.6% and 23.8% for the first and second parameter studies, respectively. The first parameter study is for a 392 megawatt reactor and the second parameter study is for a 482 megawatt reactor. (auth)
Date: August 29, 1958
Creator: Bishop, A. A. & Berringer, R.
Partner: UNT Libraries Government Documents Department

Fission and Corrosion Product Activities in Main Coolant and Atmosphere of the Vapor Container

Description: Fission products, which may leak from the fuel and activated corrosion products from stainless steel surfaces, influence the design of several systems which support the main coolant system. Since the radioactivity in the main coolant resulting from these impurities must be limited to reasonable levels during operation, a purification system becomes important. Air conditioning of the vapor container as a result of possible main coolant leakage may be necessary during and after operation to remove airborne activity. After shutdown, treatment and disposal of active effluents becomes the function of a waste disposal system. Activities are tabulated, by individual isotopes, for 10,000 hours of full power operation. (auth)
Date: May 15, 1958
Creator: Kresny, H. S. & Haga, P. B.
Partner: UNT Libraries Government Documents Department


Description: The JOFIT code fits, by a least squares technique, the curve y = A J/sub o/STAB(x-C)! from 4 to 500 points of observed data, computing the parameters A, B, C and the standard deviation of the final values of A, B, C, i.e., S/sub A/, S/ sub B/, S/sub C/. It is also possible to investigate the error in a region about the final values of A, B, C by computing the sums of the squares of the residuals at a series of points in this neighborhood. Typical computing and editing time for a 50 point problem is 2 minutes. Any size IBM-704 computer is adequate, and drums and tapes are not used. (auth)
Date: July 23, 1958
Creator: Jedruch, J.
Partner: UNT Libraries Government Documents Department

Model Study of the Pressure Drop Relationships in a Typical Fuel Rod Assembly

Description: A study was made of hydraulic characteristics of Yankee-type fuel rod assemblies using experimental and analytical methods. Two scale model fuel assemblies utilizing both ferrule and strap type arrangements were constructed and tested at atmospheric pressure and room temperature. Analytical methods using semiempirical relationships are substantiated by experimental results for both the fuel assembly having strap-type spacers and the fuel assembly having cylindrical ferruletype spacers. The experimental pressure drop across the assembly model using either straps or ferrules correlated within 5% of the value calculated by means of equations based on the equivalent diameter concept for flow inside pipes. The individual frictional drops along the rods and across the end plates and straps correlated within 15% of the predicted pressure drops. The indlvidual pressure drops across both the staggered ferrule sections and the full ferrule section correlated to within 17% of the predicted pressure drops. Comparison of the ferrule and the strap pressure drops indicates that the pressure drop across a level of straps was more than four times the pressure drop across a full ferruled section. It is concluded that the analytical methods based on the equivalent diametcr concept can be satisfactorlly used to calculate pressure drops for flow parallel to fuel rod bundles. Experimental tests on this fuel rod configuration with either straps or ferrules indicated no excessive energy losses due to vortex formations. (auth)
Date: February 1, 1959
Creator: Berringer, R. T. & Bishop, A. A.
Partner: UNT Libraries Government Documents Department

Monthly Progress Report for the Period August 1 to 31, 1958

Description: plants ln England and France. With the increasing de output of given designs and probably allow operation at higher polymer contents than orignally foreseen, thereby reducing the make-up requirements. The physical characteristics of the OMRE such as critical loading, temperature coefficient, and general stability appeared to be close to the predicted values. Radiation levels in the primary circuit area during full power operation appear to be so low that maintenance is possible during operation. The reactor has been run for a full month at 30% polymer concentration and is, at the time of this writing, brought to a still higher steady state percentage of breakdown products ln the coolant stream. No evidence whatsoever of fouling or precipitation has been observed. The reactor behaves in a routine manner in all respects and invites immediate application of the OMR principle to reactors for large scale ceniral stations. Final design on one 11.4 Mwe unit for the city of Piqua, Ohio, has now stanted. A short description is given of OMR power reactors. The use of magnetic jack mechanisms for control and safety rods provides a reactors top shield without penetrations, as well as an unpenetrated lower core vessel, still avoiding any interference from the control rods during fuel changing. The new finned-plate fuel element is mentioned as well as the use of a liquid pressurizing pump instead of nitrogen gas pressurization. It is conservatively predicted that the cost of organic liquid make-up for these designs will not contribute more than one half to one mill per kwh to the total power cost. In case operation at higher polymer concentrations appears practicable, this figure may even be lower. More detailed pricing informntion available now, has shown that the original cost estimate of around 0 per kw installed for a 150 Mwe plant can ...
Date: September 20, 1958
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department

Quarterly Progress Report for the Period July 1 to September 30, 1958

Description: An evaluation and the resulting conclusions of work performed are given for each project in which definitive progress was made. The principle progress consisted of: obtaining promising fuel fabrication results from production scale investigations of simplified UO/sub 2/ preparations methods, presintering operation elimination, increasing pellet densities and reductions of UO/sub 2/ losses; successfully brazing a ninety-five inch long experimental fuel subassembly in which all joints, including outside ferrules and control rod rubbing strips, had brazed fillets; completing the fabrication of the last group of MTR process water irradiation samples and all of the in-pile test loop samples except for the specimens containing 27% enriched pellets; performing a nuclear analysis of various fuel assembly designs as part of an over-all evaluation of fuel assembly bowing in the Yankee reactor; completing corrosion studies of primary plant materials in static autoclaves to aid in the selection of a pH control agent; developing a Compromise Design'' for the fueh assembly which incorporates a series of small design changes to eliminate the possibility of restricting control rod motion by interference due to a superposition of bowing of the fuel assembly and an adverse accummulation of mechanical tolerances; completing the calculation of the moderator temperature coefficient, the Doppler temperature coefficient, the void coefficient, neutron lifetime, and delayed neutron fraction; obtaining flux profiles, flux peaking, flux spectrum, peripheral fuel rod worth, void coefficient, temperature coefficient, and control rod worth data from the Critical Reactor Experiments with a 3: 1 water-to-uranium metal ratio core at the Westinghouse Reactor Evaluation Center; continuing instrumentation and individual component operational testing of the in-pile test loop; and analyzing the results of the post-irradiation examination of the second group of process water samples and delivery of the last group of samples to the MTR. (For preceding period see YAEC-87.) (auth)
Date: October 31, 1959
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department

A Study of Complete Loss of Coolant Flow in the Yankee Reactor

Description: The complete loss of primary coolant flow accident caused by the instantaneous loss of power to all four pumps was investigated. The principal objective was to determine the elapsed time from the initiation of the pump failure to the occurrence of bulk boiling at the outlet of the hot channel. Inherent was the determination of operating conditions necessary to eliminate the hot channel bulk boiling, these being added inertia in the primary coolant loop and scram delay time. Both the 392 Mw core and the 482 Mw core were included in the investogation. The results show that additional inertia is required as well as scram if bulk boiling is to be eliminated. As a result of several of the assumptions made, the investigation was limited to a study of that region of the reactor core where heat transfer takes place between the fuel and the coolant; primary system components such as the pressurizer and steam generator were excluded. The film heat transfer coefficient was varied as a function of flow rate but not as a function of coolant temperature, thus boiling phenomena were not included. Distributed parameter effects were incorporated in equations which describe the core thermal kinetics by an approximation technique. An approximation technique is also presented which allows the extrapolation of the results to the two pump failure case. Data for the investigation were obtained by the use of a general purpose d-c analog computer. (auth)
Date: November 1, 1958
Creator: Gallagher, J. M., Jr. & Hunter, D.
Partner: UNT Libraries Government Documents Department

A Study of Decontamination Agents for Use in the Yankee Reactor

Description: A study of chemical decontamination agents for possible use in the Yankee Reactor is presented. The necessity for chemical decontamination is discussed along with the properties sought in an ideal cleaning solution. A survey of available decontamination information is presented together with a description of the experimental test progam carried out at the Westinghouse Atomic Power Department in an effort to develop new and improved chemical methods for the removal of oxide scals and radioactive contamination from Yankee reactor primary coolant system components. On the basis of bench scale testing of various cleanup procedures, a method employing basic permanganate and citrate solutions was selected as offering the most promise for successful solution of Yankee primary loop contamination problems. Data are presented concerning the attack rates produced by these reagents on various reactor materials of construction, as is an evaluation of the possibility of caustic embrittlement of stainless steel resulting from the use of the basic permanganate solution. A discussion of future larger scale testing of the proposed decontaminants is also included. (auth)
Date: November 1, 1958
Creator: Watkins, R. M.
Partner: UNT Libraries Government Documents Department

Thermal Deflection of the Yankee Fuel Assembly From Linear and Non-Linear Temperature Gradients

Description: Theoretical conditions were investigated for determining the thermal deflection of the definitive design unshrouded Yankee fuel assembly and the compromise design fuel assembly. In addition, an experimental study was completed on the thermal deflection of the compromise design fuel assembly from linear and nonlinear temperature gradients. All of the theoretical analyses performed on the definitive design clearly indicate that excessive bowing in the order of 0.150 inch may occur within the reactor during normal operation. Slnce the nominal clearance available in this design between a fuel assembly and an adjacent control rod is about 0.120 inch, interference between a fuel assembly and its adjacent control rod can occur, resulting in restricting the movement of the control rod. When buildup of mechanical tolerances are considered, this condition becomes more severe. Redesign of the definitive design fuel assembly thus became mandatory to prevent interference with the movement of control rods. The compromise fuel assembly design evolved as a result of these considerations. The compromise design employs the use of the brazed ferrule fuel sub-assembly design concept, but incorporates several fundamental changes in design that increase clearances adjacent to the control rod to reduce thermal bowing of the complete assembly. The compromise fuel assembly is allowed to deflect under thermal distortion to a limited extent and is designed so that under the worst set of circumstances the available clearance between the fuel assembly and control rod is at least 0.001 inch. The compromise design therefore, constitutes a safe design from a thermal deflection standpoint. (auth)
Date: March 1, 1959
Creator: Johnson, C. G.
Partner: UNT Libraries Government Documents Department

Thermoelectric Nuclear Fuel Element. Addendum to Final Report

Description: A twenty-couple fission-heated thermoelectric generator, TE-5, was tested in the Westinghouse Testing Reactor. The open circuit voltage, instantaneous closed circuit voltage, steady state closed circuit voltage, instantaneous closed circuit current, and steady state closed circuit current were monitored during the 23-day cycle. The initial power output was 110 watts (48 amps at 2.3 volts) at a TH of 660 to 690 deg C and a T/sup C/ of 130 deg C. After 23 days the output was approximately one watt. (auth)
Date: March 30, 1962
Creator: Danko, J. C.
Partner: UNT Libraries Government Documents Department

Thermoelectric Nuclear Fuel Element Quarterly Progress Report, April-June 1961

Description: Uranium-bearing thermoelectric compounds are now being prepared by tantalum bomb melting and by the hydride process. Tests of devices made up from these compounds indicate that the main fabrication problems are densification and contact bonding. Data from a hot-swaged pellet and a swaged device of US/sub 2/ indicate some promise for that compound. Improvements in techniques of thermoelectric parameter measurements include programming of automatic test data recording at desired intervals around the clock; increased accuracy and versatility of measurements through use of a newly-constructed adjustable precision resistor; and a method for measuring which should lead to an experimental means for determining the thermoelectric figure of merit, Z. Potential profile studies on PbTe pelleta are yielding important information on contact resistance parameters. A fission-fired thermoelectric generator is being prepared for the next in-pile test. (auth)
Date: July 10, 1961
Partner: UNT Libraries Government Documents Department

Thermoelectric Nuclear Fuel Element Quarterly Progress Report for April- June 1960

Description: A hot-pressing system was designed and is adaptable for bomb melting experiments up to 2000 deg C. Two devices were constructed for measuring. One was designed to measure Seebeck coefficients, resistivities, and bond resistances of swaged and machined thermoelectric wafers and the other was designed to measure thermoelectric parameters of cylindrical thermoelectric pellets up to 1000 deg C. Design studies on a fission-fired thermoelectric generator were completed. Seebeck coefficient and resistivity were determined for Li/sub .06/Ni/ sub .94/ and p type PbTe as a function of thermal fiux at 400 deg C (average). Designs of prototype thermonuclear fuel elements are presented which include single-leg, double-leg, multi-junction notched-bar, and multi-junction pelletized designs. The effects of heattreatment on the thermoelectric properties of n and p type PbTe were determined at 68 deg F. The compatibility of PbTe and GeTe with cladding materials was investigated at 510, 600, and 650 deg C. The results of life testing a swaged single-leg GeTe element are discussed. The nuclear characteristics of a 500-kw thermoelectric core that employs a rod-type fuel element were calculated. Designs of a thermoelectric reactor system that utilizes thermal circulation are presented. (For preceding period see WCAP- 1545.) (C.J.G.)
Date: July 10, 1960
Creator: Blankenship, W. P.; Goodspeed, R. C.; Markley, R. A. & Mitchell, P. V.
Partner: UNT Libraries Government Documents Department

Two Dimensional Diffusion Theory Studies of Control Rod Worths and Flux Peaking in the Yankee Reactor

Description: BS> Procedures are discussed for the utilization of a two dimensional diffusion theory computer program (QED) in the calculation of neutron flux peaking and control rod worths. Some typical results obtained for the first Yankee core are presented. Methods for reducing large problems to a size acceptable to the computer program are described. (auth)
Date: August 1, 1958
Creator: Minton, G. H. & Tirellis, G.
Partner: UNT Libraries Government Documents Department

Bulk Biological Shielding Aspects of the Yankee Core

Description: The analysis of the radiation sources within the primary concrete shielding of the Yankee Atomic Electric Plant is summarized The dose due to these sources at a point on a transverse centerline and outside the plant container was calculated. The dose at two points on the axial centerline was found as well as the extent to which the water is actiwated during its passage through the pressure vessel. (auth)
Date: October 1, 1957
Creator: Graves, H. W., Jr.; Eich, W. J. & Williams, H. T., Jr.
Partner: UNT Libraries Government Documents Department