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Program for the Development of Plutonium Recycle for Use in Light-Water- Moderated Reactors. Eleventh Quarterly Report, October 1-December 31, 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. Autoradiographs of unirradiated fuel indicated approximates 10% microsegregation of the PuO/sub 2/. The irradiations of flux wires and the project fuel element in VBWR are described; the fuel element was removed after approximates 5000 Mwd/t … more
Date: January 15, 1964
Creator: Robkin, M. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Environmental Testing of a B4C-Ni 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 examina… more
Date: October 15, 1963
Creator: Megerth, F. H. & Zimmerman, D. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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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 tot… more
Date: July 15, 1963
Creator: Robkin, M. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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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 app… more
Date: July 15, 1963
Creator: Garelis, Edward & Meyer, P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Grain Growth of UO2. Part I

Description: Abstract: (1) Grain growth in UO2 pellets was studies between 100 C and 2600 C. The pellets were encapsulated in small vacuum-tight tungsten containers in an argon atmosphere. (2) The grain size-time relationship could be expressed by an equation. A low exponent, m>_ 1/3, was found in those experiments and is related to the type of UO2 investigated. An activation energy of 65 kcal/mole was obtained for the grain growth process. The time exponent, m, increased with increasing temperature if the … more
Date: August 15, 1963
Creator: Hausner, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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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 o… more
Date: September 20, 1963
Creator: Reynolds, M. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Maritime Loop Irradiation Program for Savannah I Fuel Post-Irradiation Examination of SI5BM Fuel Assembly

Description: Abstract: A stainless steel clad 9-rod assembly fabricated by The Babcock & Wilcox Company was irradiated in a boiling water loop of the General Electric Test Reactor. A post-irradiation examination revealed no significant dimensional changes on the fuel rods. the results of mass spectrometric analysis made of the pelletized UO2 fuel indicated a maximum burnup of 11,500 MWD/tonne was attained by Rod B-4 during the exposure.An x-ray diffraction examination of an unirradiated fuel sample revealed… more
Date: November 7, 1963
Creator: Mathay, P. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fuel Failure Examinations and Analyses in the High Power Density Program

Description: Summary: The High Power Density Project includes a comprehensive fuel development program which has the objective of developing and demonstrating the performance of a nuclear reactor core having a high power density, long fuel life, and low fabrication cost. The fuel program is made up of two principal tasks. Task 1A consists of irradiation tests in the VBWR of Type 304 stainless steel clad, UO2 pellet type fuel rods fabricated by current commercial processes. Task 1B consists of the investigat… more
Date: September 16, 1963
Creator: Arlt, W. H. & Vandenberg, S. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Specific Zirconium Alloy Design Program Quarterly Progress Report: Sixth Quarter, July - September, 1963

Description: Summary: Fundamental studies in support of the alloy design work are complete except for the experimental determination of the diffusion of oxygen in alloy-doped non-stoichiometric ZrO2. Over 100 oxidation runs have now been made on samples of ZrO2 doped with 1 mole percent of the oxides of Al, Y, Fe, Cr, and Ni. The first round testing of 31 alloys is now essentially complete. Analysis of the steam corrosion rate and hydriding raw data taken at 300, 400, and 500 degrees C indicates that Zr-C… more
Date: October 1, 1963
Creator: Klepfer, H. H.; Jaech, John L.; Douglass, D. L. (David Leslie), 1931-; Blood, R. E. & Perrine, H. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Sixth Quarterly Progress Report, July 1-September 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 app… more
Date: October 15, 1963
Creator: Garelis, Edward & Meyer, P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A Controlled-Environment Steam Corrosion Facility

Description: Abstract; Technical report describing a low-flow autoclave system developed for out-of-pile corrosion testing of materials in controlled environment steam up to 500 C. The system has been set up in triplicate to provide for the exposure of various zirconium alloys to steam at 300, 400, and 500 C. The oxygen and hydrogen of the steam were controlled at 25 ppm and 3 ppm, respectively, to simulate the gas conditions from radiolytic water decomposition found in a boiling water reactor. The autocla… more
Date: October 1963
Creator: Nelson, W. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Uranium Dioxide Fuel Rod Center Melting Test in the Vallecitos Boiling Water Reactor

Description: Technical report describing that as part of the AEC Fuel Cycle Program, tests are being conducted to evaluate the significance of current fuel design limitations that do not permit the maximum fuel temperature to exceed the melting point of UO2. The reliability of prediction of the fuel rod operating conditions that will cause melting of the UO2 was evaluated by means of a calibration test conducted in the Vallecitos Boiling Water Reactor. Conclusions: (a) The central portion of the 3.15-cm dia… more
Date: November 15, 1963
Creator: Williamson, H. E. & Hoffmann, J. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Accurate Nuclear Fuel Burnup Analyses; Eighth Quarterly Progress Report, (September - November 1963)

Description: The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Date: December 1, 1963
Creator: Rider, B. F.; Ruiz, C. P.; Luke, P. S., Jr.; Peterson, J. P., Jr. & Smith, F. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel

Description: Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was inc… more
Date: March 1964
Creator: Ogawa, S. Y.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Localized Corrosion of Stainless Steels and High-Nickel Alloys in Simulated Superheat Reactor Environment

Description: Abstract. A program was instituted to study and reproduce the in-reactor intergranular failures of Type-304 stainless steel fuel cladding found in superheated steam. The program was directed toward finding ways to eliminate the cause of failure or to use improved alloys that would be less susceptible to failure. A materials screening test was developed in the out-of-pile superheat facilities with 1.5 ppm chloride added as sodium chloride to the recirculating water in the presence of typical … more
Date: February 1964
Creator: Pearl, W. L.; Gaul, G. G. & Wozadlo, G. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Statistical Techniques Used in the Specific Zirconium Alloy Design Program

Description: Technical report describing the statistically designed empirical approach being used to choose a candidate Zr alloy optimum with respect to corrosion and hydriding rates in steam and having acceptable mechanical properties. The statistical techniques used and the reasons for their use are discussed in detail, with emphasis on the estimation of corrosion rates. Estimation of response surfaces is also considered.
Date: November 11, 1963
Creator: Jaech, John L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Oxidation Mechanism of Zirconium and Its Alloys. [Part] II. Oxide Plasticity

Description: Abstract: The question of how crack-free, protective oxide films can form on zirconium during oxidation when the Pilling-Bedworth ratio is about 1.5 has been considered by a study of the relative plasticity of various forms of zirconia. Hot hardness measurements showed that doping mono-clinic zirconia with iron, nickel, or chromium resulted in softer (more plastic) structures and that yttrium additions slightly reduced the plasticity. Calcia-stabilized cubic zirconia was found to be more pla… more
Date: February 20, 1964
Creator: Douglass, D. L. (David Leslie), 1931-
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: October 1 - December 31, 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. Quarterly progress: Forty-six fuel pellet faces have been auto-radiographed. These faces have been prepared from twenty-three pellets by making an exposure before and after the removal of an additional ten mils of fuel. A substantial number… more
Date: January 15, 1964
Creator: Robkin, M. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Specific Zirconium Alloy Design Program Quarterly Progress Report: Seventh Quarter, October - December, 1963

Description: Summary: All experimental work under the Corrosion Mechanism task has been completed. The remaining topical reports are being prepared by D. L. Douglass, now on assignment at Mol. Experimental work on the first round of 31 alloys and on the second round of 10 alloys has been completed. Steam exposures of at least 3000 hours were finished for all the alloys at all test temperatures, with exposures of some coupons to 6700 hours. Mathematical expression have been derived to describe all first r… more
Date: January 1964
Creator: Klepfer, H. H.; Jaech, John L.; Blood, R. E.; Perrine, H. E. & Urata, M. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

General Corrosion of Incoloy-800 in Simulated Superheat Reactor Environment

Description: The 300 series stainless steels had been selected originally as the reference fuel cladding material for utilization in several superheat reactor (SHR) systems being built as part of the United States Atomic Energy Commission (USAEC) program. The adequacy of the general corrosion resistance of Type-304 stainless steel for superheat fuel cladding was confirmed in the Phase I portion of the study reported previously. Fuel jacket failures that occurred in Type-304 stainless clad fuel elements exp… more
Date: March 1964
Creator: Pearl, W. L.; Brush, E. G.; Gaul, G. G. & Wozadlo, G. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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