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[Alice Faye filming State Fair]

Description: Photograph of famous actress and singer Alice Faye, standing onset as character Melissa Frake in the 1962 film State Fair. State Fair was shot in September and October 1961 at the Texas State Fair Grounds and at the Oklahoma City State Fair Grounds. Puerto Rican actor and film director José Ferrer is seen seated in his chair while the film crew with the camera and boom are visible on the right-side of the photograph.
Date: [1961-09,1961-10]
Creator: de Bruchard, Georgette
Partner: UNT Libraries Special Collections

[Alice Faye filming scene for State Fair]

Description: Photograph of famous actress and singer Alice Faye, standing onset as character Melissa Frake in the 1962 film State Fair. State Fair was shot in September and October 1961 at the Texas State Fair Grounds and at the Oklahoma City State Fair Grounds. Puerto Rican actor and film director José Ferrer is seen seated in his chair while the film crew runs around getting the set ready, the camera and boom being visible on the right-side of the photograph.
Date: [1961-09,1961-10]
Creator: de Bruchard, Georgette
Partner: UNT Libraries Special Collections

[Photograph of José Ferrer and Alice Faye]

Description: Photograph of famous actress and singer Alice Faye, standing onset as character Melissa Frake in the 1962 film State Fair. State Fair was shot in September and October 1961 at the Texas State Fair Grounds and at the Oklahoma City State Fair Grounds. Puerto Rican actor and film director José Ferrer is standing opposite of Faye and conversing with her.
Date: [1961-09,1961-10]
Creator: de Bruchard, Georgette
Partner: UNT Libraries Special Collections

[Portrait of Alice Faye]

Description: Photograph of famous actress and singer Alice Faye, standing onset as character Melissa Frake in the 1962 film State Fair. State Fair was shot in September and October 1961 at the Texas State Fair Grounds and at the Oklahoma City State Fair Grounds.
Date: [1961-09,1961-10]
Creator: de Bruchard, Georgette
Partner: UNT Libraries Special Collections
open access

Hazards Evaluation of the SM-1 Penetrated Gasket

Description: Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.
Date: September 8, 1961
Creator: Coombe, J. R.; Gebhardt, F. G. & James, B.
Partner: UNT Libraries Government Documents Department
open access

Hazards Report for Insertion of the PM-1-M-2 Element in the SM-1 Core II

Description: Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as defined by Alco Products, Inc., it is recommended that if time permits that critical facility design verification be accomplished. The PM-1-M-2 test element meets the minimum requirements for insertion in SM-1 Core II and in view of the importance of this element to the PM-1 and PM-3A program, should be considered for insertion.
Date: September 1, 1961
Creator: Coombe, J. R.; Scoles, J. F.; Brondel, J. O. & Lee, D. H.
Partner: UNT Libraries Government Documents Department
open access

Consolidated Edison Thorium Reactor: Core Assembly Physics and Zero Power Tests Report

Description: Results of zero power tests to confirm the nuclear design of the Consolidated Edison Thorium Reactor Core are presented. The maximum core reactivity, control rod worths, and power distribution were measured.
Date: September 1961
Creator: Edlund, Milton C. (Milton Carl), 1924-; Ball, R. M.; Deddens, J. C. & Flickinger, R. F.
Partner: UNT Libraries Government Documents Department
open access

Steady State and Transient Thermal and Hydraulic Analysis of SM-2 Termination Report

Description: Abstract: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during a loss of flow accident, ART-02, a one-dimensional code, was used. This analysis indicates the SM-2 core is safe from burnout under steady state operation at design power level (28 tMW) because (1) no nucleate boiling exists, and (2) the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of flow transient because the minimum burnout ratio in the hottest element channel of 1. 82 is above the minimum design criteria of 1. 5.
Date: September 8, 1961
Creator: Segalman, I. & Bradley, P. L.
Partner: UNT Libraries Government Documents Department
open access

Design Criteria for Irradiated Vessels Task 6.0 Summary Report

Description: Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the criteria are included.
Date: September 29, 1961
Creator: McLaughlin, D. W.
Partner: UNT Libraries Government Documents Department
open access

Investigations of Ice-Free Sites for Aircraft Landings in East Greenland, 1959

Description: Abstract: "Thirty-three specific landing sites were investigated in the ice-free land area of East Greenland between Scores by Sund and Loch Fyne. Eight of these are considered suitable for emergency landings in summer by heavy cargo planes, and several more for light cargo planes. Several sites were investigated for the Royal Greenland Trade Department in the Scorebysund - Kap Tobin area. A 1550-foot airstrip was located on a gravel terrace in the Jaettedal, eight miles northwest of Kap Tobin, and a short strip requiring some construction work was located near Kap Tobin. An 11,500-foot airstrip was tentatively laid out on a gravel terrace at Storelv, near Moskusoksefjord. Utilization of several of these sites can add a significant safety factor to commercial or military aircraft operations in East Greenland. Reconnaissance observations verify the presence of abundant emergency sources of fresh water in East Greenland; analyses of 36 samples indicate water of good to excellent chemical quality."
Date: September 1961
Creator: Hartshorn, Joseph Harold; Stoertz, George E.; Kover, Allan N. & Davis, Stanley N.
Partner: UNT Libraries Government Documents Department
open access

Engineering bases for power levels and exposures, October 1961--December 1962

Description: It is the purpose of this document to provide assistance to the Manufacturing Section personnel in determining their future operating plans. In general, the inter-relationship of such engineering parameters as projected flow rates, reactor orificing pattern, fuel element performance, and process limits have been considered. Based on these engineering parameters and related process economics, suggested reactor ``Operating Plans`` are graphically presented in this document. It is emphasized that these plans do not reflect operational considerations which may modify the desirability of the indicated power level increase nor has allowance been made for major projects, major maintenance outages, etc. Many factors which only manufacturing personnel are capable of evaluating may make it desirable to operate above or below these operating plants. These plans are designed to present reasonably achievable but perhaps optimistic power levels together with process limits which will be approached or will possibly limit reactor power levels unless limit revisions can be effected. It should be noted that the engineering parameters and basic assumptions which have been factored into these plans are subject to continual re-evaluation and revision. In a strict sense, these plans are out-dated even as they are published. However, their value will lie primarily in illustrating the approximate conditions under which each reactor can be operated during CY-1962; thus, permitting some perspective of individual and collective reactor operation.
Date: September 18, 1961
Partner: UNT Libraries Government Documents Department
open access

Numerical results of production test IP-326-I, low flow calibration tests at DR and C reactors, and of production test IP-395-I, removal of high tank discharge line strainer gates at B, D, and F reactors

Description: Low flow calibration tests have now been completed at all old reactors as a part of the program to determine the adequacy of the old reactor last ditch water backup system more accurately. In addition, the high tank discharge line strainer gates have been removed at B, D, and F reactors in order to increase the emergency high tank flow rates to the reactors and thereby increase high tank water backup adequacy. The purpose of this document is to list the results of the last two low flow calibration tests at DR and C reactors (calibration test results at B, D, F, and H reactors are listed in reference 7) and to list the increased high tank flow rates that were obtained at B, D, and F reactors when the strainer gates were removed.
Date: September 29, 1961
Creator: Benson, J. L.
Partner: UNT Libraries Government Documents Department
open access

NPR Reactor shield calculations

Description: At the request of IPD Personnel, calculations on neutron and gamma attenuation were made for the NPR shield. The calculations were made using a new shielding computer code developed for the IBM 7090. The calculations show the thermal neutron flux, total neutron dose rate, and gamma dose rate distribution through the entire shield assembly. The calculations show that the side and top primary shield design is adequate to reduce the radiation level below design tolerances. The radiation leakage through the front shield was higher than the design tolerances. Two alternate biological shield materials were studied for use on the front face. These two materials were iron serpentine concrete mixtures with densities of 245 lb/ft{sup 3} and 265 lb/ft{sup 3} (designated by I-S-245-P and I-S-265-P, respectively). Both of these concretes reduced the radiation below design tolerances. It is recommended that the present front face biological shield be changed from I-S-220-P to I-S-245-P. With this change the NPR shield is adequate according to these calculations. The calculations reported here do not include leakage through penetration in the shield.
Date: September 27, 1961
Creator: Peterson, E. G.
Partner: UNT Libraries Government Documents Department
open access

Reactivity and graphite temperature effects of a helium-nitrogen atmosphere. Final report of PT IP-358-AC

Description: The purpose of the production test was to {open_quotes}investigate the reactivity and temperature effects and the associated operating problems resulting from the use of nitrogen instead of carbon dioxide as a constituent of the K Reactor atmosphere.{close_quotes} This report summarizes the reactivity and reactivity-associated graphite temperature effects observed during the first four months of the test which was initiated at KE Reactor on December 5, 1960.
Date: September 29, 1961
Creator: Bailey, G. F.
Partner: UNT Libraries Government Documents Department
open access

TOE work sheets

Description: This report provides operational worksheets detailing charge/discharge, maintenance, fuel element ruptures, leaks, tube replacements, and production data.
Date: September 14, 1961
Creator: Chatten, J. C. L.
Partner: UNT Libraries Government Documents Department
open access

Irradiation effects in hot-pressed fuel elements

Description: Two Hanford hot-pressed fuel elements were irradiated in the Materials Test Reactor to an approximate exposure of 1000 MWD/T (0. 11 a/o burnup or 5. 8 {times} 10{sup 19} fissions/CM{sup 3}). One element represented the best bond quality, the other element represented a poor bond due to unoutgassed uranium. After irradiation, radiometallurgical examination revealed extensive bond separation in the unoutgassed fuel element. The Ni{sub 2}Al{sub 3} layer thickened slightly and cracks were observed in this layer throughout both elements. Any cracks observed in the uranium terminated at the nickel layer. The double-worked diffusion closure showed no sign of corrosion, pitting or intergranular attack.
Date: September 1, 1961
Creator: Tverberg, J. C.
Partner: UNT Libraries Government Documents Department
open access

Fabrication Costs

Description: This document from the principal engineer of the manufacturing section at Hanford calls attention to irreconcilable costs for two fuel fabrication studies. Specifically, the major trouble areas appear to be in refinery, green salt and machining steps.
Date: September 13, 1961
Creator: Lang, L. W.
Partner: UNT Libraries Government Documents Department
open access

The Preparation and Irradiation Behavior of Chemically-Nickel Plated Aluminum-Jacketed Fuel Elements

Description: Nickel plated aluminum was considered as a jacketing material for nuclear fuel elements as early as 1954, and both static and dynamic corrosion tests were carried out by Argonne National Laboratories and by Atomic Energy of Canada Ltd., employing demineralized water at temperatures of from 260 to 316{degree}C. Results generally indicated that the nickel had excellent corrosion resistance; however, difficulties were experienced in achieving satisfactory continuity and adhesion of the plate; subsequent work emphasized Ni-Aluminum alloy development. At Hanford, our earliest experience employed Ni plate on aluminum-jacketed fuel elements primarily to minimize mechanical damage to the jacket surface during an irradiation test. The appearance of these fuel elements after discharge suggested that the nickel plate might also be a highly satisfactory coating for corrosion and abrasion resistance. Incentives were manifold, including reducing the incidence of in-reactor fuel element failures and permitting reduction of the aluminum jacket thickness with a concomitant increase in space available for uranium or for cooling water passage. A program has been carried out for the past three years aimed at determining various methods of employing nickel plated aluminum jacket material and testing the capabilities of high quality commercially adequate plate. Almost exclusively chemically deposited plate has been tested, primarily because of the excellent plate-thickness control inherent in the process compared to electrodepositon.
Date: September 5, 1961
Creator: Jacky, G. F.
Partner: UNT Libraries Government Documents Department
open access

B and D downcomer prototype data

Description: Available pressure data on the B and D downcomers and approach piping is tabulated herein. All readings were taken with an air bubbler system; readings given in psi were taken on a Bourdon-type pressure-vacuum gauge, readings given in inches of water were read on an oil-filled manometer. In each case, fluctuations occurred and the mean value was estimated by eye. Data are presented as a function of flow rate and temperature.
Date: September 22, 1961
Creator: Corley, J. P.
Partner: UNT Libraries Government Documents Department
open access

Power calculation accuracy with dual downcomer operation

Description: The C reactor is presently, operating with all process water discharging through one downcomer. Pressure measurements at the downcomer lid show the top tray is nearly flooded, thus prohibiting further flow increases until the downcomer is modified or both downcomers are used simultaneously. Operation with two downcomers can result in two sources of error in the power calculation: (1) Errors in measured outlet temperature can occur, apparently due to the increased venting capacity. The magnitude of the effect is probably unpredictable and could not be compensated for in the power calculation. Since the temperature error occurs only downstream of the downcomer vents, the logical solution is to move the outlet temperature devices to the risers. The method of administering bulk temperature limits will have to be changed to account for the thermal shield flow entering downstream of the outlet thermohms. (2) Simultaneous flow and temperature unbalances can result in power calculation errors since there is no way of knowing the actual flow through each downcomer. It is assumed that the power calculation would have to be made from the total flow and the average of the two outlet temperatures. The primary purpose of this study was to determine whether the power calculation errors due to flow and temperature unbalances would be significant.
Date: September 11, 1961
Creator: Renberger, D. L.
Partner: UNT Libraries Government Documents Department
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