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A Program of Two-Phase Flow Investigation Quarterly Report: Second Quarterly Report, July-September, 1963

Description: Task A: Modification and Preparation of Experimental Facility. With the exception of the insulation of modified components, the experimental facility is complete. Insulation will be completed by the end of September. the system has been charged with Refrigerant-22 and preliminary loop performance tests have been completed without the test sections.. Task B: Design and Construction of Test Sections. The stainless steel test section has been prepared complete with end flanges and pressure tap locations. Wall thickness tolerances have been ultrasonically checked. Test section inlet and discharge assemblies are being completed and the whole assembly will be ready for installation by the end of September. Glass sections from the same drawn length which will make up the final test sections have been received for pressure tests. The final coated sections and the associated inlet and discharge fittings will be ready for assembly by the end of September. The above sections were ordered after complete preliminary tests defined the properties required of these test sections. Task C: Design and Construction of Test Stand: The mechanical design and drafting of the structural components and drive system is complete. The electrical control system for the platform orientation has been constructed and two modes of operation have received preliminary checkout on the breadboard circuit. Machining of the supported plates and assemblies and the main structural assembly of the stand are complete. Assembly of the platform and electrical components will be completed by the end of September and final alignment and adjustments will be caries out during October. Task E: Pressure and Temperature Instrumentation for Test Section: The temperature transducers have been ordered and their delivery is scheduled for the third week of September. The bridge circuits have been designed and their fabrication is in process. The preamplifiers and the power supply for the temperature transducers ...
Date: September 23, 1963
Creator: Staub, F. W. & Zuber, N.
Partner: UNT Libraries Government Documents Department

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-Cr and Zr-Cu-Fe alloys show the most promise for development for service in steam over the entire temperature range 300-500 degrees C. Maximum resistance to corrosion hydrogen embrittlement requires high initial ductility and thus low, perhaps less than ~2.5 a/o total alloy content. For any composition, susceptibility to hydrogen embrittlement depends on crystallographic texture of the component; under certain circumstances hydrogen embrittlement may be high anisotropic. The second-round testing of 10 selected Zr-Cr and Zr-Cu base alloys is now about 50% complete. Three alternate fabrication schedules were evaluated; and the preliminary results indicate that the Zr-Cu alloy tested is less sensitive to heat treatment than is the Zr-Cr alloy tested. Raising the final alpha annealing temperature from 565 degrees C to 788 degrees C gives better over-all corrosion and hydrating performance for both the Zr-Cr and Zr-Cu alloy tested. Beryllium additions to Zr-Cr or Zr-Cu do not appear to be advantageous. Nickel additions to Zr-Cu do not give an over-all improvement. Nickel additions to Zr-Cu give about the same improvement over Zr-Cu as did iron additions to the Zr-Cu in the first-round test.
Date: October 1, 1963
Creator: Klepfer, H. H.; Jaech, John L.; Douglass, D. L. (David Leslie), 1931-; Blood, R. E. & Perrine, H. E.
Partner: UNT Libraries Government Documents Department

Transition Boiling Heat Transfer Program; Third Quarterly Progress Report, July - September 1963

Description: Summary: Initial critical heat flux, transition boiling temperature fluctuation, and film boiling coefficient data have been obtained on a two-rod cluster assembly at 1000 psia and 25 to 90 percent steam qualities. A representation showing the range of critical heat flux data is presented. Typical temperature recordings which indicate transition and film boiling behavior are shown. Fabrication of a new high pressure observational test section is nearly complete. An optical table and illumination system has been build and operationally tested for photographic use on the new observational section.
Date: October 1, 1963
Creator: Quinn, E. P.
Partner: UNT Libraries Government Documents Department

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 autoclave internals were so designed to result in a temperature variation between specimens under test of less than 2C.
Date: October 1963
Creator: Nelson, W. B.
Partner: UNT Libraries Government Documents Department

Fuel Cycle Program Design and Fabrication of Special Assembly 10-L : Compacted Powder Fuel Rods Clad With 0.127-MM Wall Stainless Steel

Description: Technical report describing sixteen fuel rods clad with thin type 304 stainless steel and filled with vibratory compact powder UO2 that were fabricated and incorporated into a bundle for irradiation testing in the VBWR. The UO2 powders were tested for gas content. N2, CO, and H2 were the principal gases evolved by both type of UO2, but the arc-fused UO2 released about ten times as much gas as the Dyna Pak UO2. The amount of gas released was also a function of particle size and temperature. The gas evolution data were used to design the gas plenum to accommodate the absorbed gases along with the fission gases.
Date: October 1963
Creator: Ogawa, S. Y. & Williamson, N. E.
Partner: UNT Libraries Government Documents Department

High Power Density Development Project: Fourteenth Quarterly Progress Report, July-September 1963

Description: Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC. Five tasks are in progress: (1) Task 1A-High Power Density Fuel Development. The number of assemblies has been reduced to seven as a result of the failure of two pellet fuel assemblies. The average burnup of the group operating as of September 1 is 7500 MWD/T. (2) Task 1B-Fuel Fabrication Development. Assembly. Assembly 12S gave positive signals of being a leaker under the multi-type in-core sampler and was declared failed based on the in-core results and visual observation of a cracked rod. Modifications to the instrumented fuel assembly probes were made by removing the failed flow meter rotors to allow continued use of the flux detectors and thermocouples. Flux detectors and thermocouples performed properly after reactor start up. Flux wire tubes were found to be kinked such that their use was prohibited. (3) Task II-Stability, Heat Transfer and Fluid Flow. A series of noise recordings of fluxes, flows, and temperatures has been made at 91 MWt at the Big Rock Point plant. Preliminary analyses of some of the these records were made to obtain noise amplitude as a function of frequency. Thermocouple response tests were performed to verify the temperature measurement obtained during the steady-state noise tests at Big Rock. (4) Task III-Physics Development. Plans for achieving optimum performance from the Big Rock plant are being based on the concept of maintaining a fixed power shape throughout each operating cycle. The desired shape for the present cycle has been computed. Methods of selecting control rod patterns to maintain this shape are being investigated for use in the on-line computer. The computer was put on line during plant startup in August, and is presently performing ...
Date: October 1963
Creator: Holladay, R. L.
Partner: UNT Libraries Government Documents Department

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 diameter uranium dioxide fuel column melted. It appears that the UO2 was molten to a radius of 1.22 cm in the peak power region. The maximum extent of melting probably occurred during the peak power run when the kdT in this region of the rod reached 171 watts cm. The estimated radius of melting from metallographic examination indicates the kdT for sintered UO2 is 89 watts/cm. This supports a calculated estimate for sintered UO2 thermal conductivity published by D. R. deHalas and G. R. Horn. The results of the previous calibration run and subsequent experimental data by Lyons are also consistent with the value. This conclusion is contingent on the interpretation of the post-irradiation crystal structure of the UO2. Insufficient data are available on the mechanisms by which various UO2 crystal structures are formed to permit a positive identification of the extent of melting and correlation with time of operation. No conclusion can be drawn as to whether the thermal conductivity of the UO2 changed with operation. (b) Although extensive UO2 melting occurred, there was no indication of fuel rod clad failure. (c) Axial heat transfer by convection in the molten UO2 was significant.
Date: November 15, 1963
Creator: Williamson, H. E. & Hoffmann, J. P.
Partner: UNT Libraries Government Documents Department

Design and Fabrication of Pellet Fuel Rods Clad With Thin Wall Stainless Steel

Description: Summary: Stainless steel clad nuclear fuel cycle costs can be reduced to those associated with Zircaloy clad fuel or potentially lower by reducing the thickness of the clad tube wall until performance penalties offset the savings associated with the reduction in parasitic neutron absorption. To demonstrate the feasibility and investigate performance capabilities of thin clad fuel rods for power reactor application an assembly was fabricated with 0.0127 cm (5 mil) thick stainless steel cladding tubes for irradiation testing in the Vallecitos Boiling Water Reactor (VBWR). The fuel bundle was placed in the VBWR and irradiation was begun in November, 1961. The irradiation is scheduled to continue until the target exposure of 2.74 x 10(20) fissions/cc (10,000 MWD/T of uranium) average burnup is reached. Destructive examinations of fuel rods will be performed at regular intervals throughout life to determine fuel rod performance.
Date: February 1964
Creator: Hoffmann, J. P.
Partner: UNT Libraries Government Documents Department

Prediction of Two-Phase Critical Flow Rate

Description: Technical report of a proposal of an analytical model to predict two-phase critical flow rate. The model is based upon thermal equilibrium, a "lumped" treatment of the two-phase velocity (each phase is represented by a single mean velocity), and upon the neglect of frictional and hydrostatic pressure losses. A comparison, of the proposed predictions with available test results and previous analyses shows that: (1) The present model agrees very well with the published test data. (2) In contrast to all other analyses, the model requires no assumption about the gas void fraction.
Date: October 1963
Creator: Levy, S.
Partner: UNT Libraries Government Documents Department

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 incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter of the fuel rods is 1.308 cm (0.515 inch), and the clad wall thickness if 0.051 cm (0.020 inch). the cladding material is Type-304 stainless steel. The fuel pellets were all centerless ground to achieve a uniform outside diameter and thereby control the pellet-to-clad diametral clearance within a range of 0.076 to 0.102 mm (0.003 to 0.004 inch). Operation of the fuel rods will be at high specific power levels with surface heat fluxes of about 157 W/cm(2) (~500,000 Btu/h-ft(2)). The assembly was designed for a lifetime of 4.1 x 10(20) fission/cc (15,000 MWD/T) exposure.
Date: March 1964
Creator: Ogawa, S. Y.
Partner: UNT Libraries Government Documents Department

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.
Partner: UNT Libraries Government Documents Department

Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Ninth Quarterly Report, October-December 1963

Description: Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, fuel testing in TREAT, fuel performance evaluation, fast-flux irradiation of fuel, and reactor dynamics and design.
Date: January 1964
Creator: Leitz, F. J.
Partner: UNT Libraries Government Documents Department

Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: July 1 - September 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. Quarterly progress: The densitometry procedure (for measurement of alpha autoradiographs of fuel pellets) has been modified to eliminate the need for a second emulsion. The existence of a problem of latent image fading and non-reciprocity of the high-resolution emulsion has been recognized. A tentative procedure has been worked out to correct these emulsion difficulties. the number of polished pellets has been increased to thirteen. The number of hot spots per pellet has not changed appreciably. The largest spot seen is irregular with an estimated volume equivalent to that of a sphere of 35 mil diameter with a PuO2 concentration in the neighborhood of 60%. The VBWR irradiation run now under way is not scheduled to end until October. To the end of the last run the cumulative exposure reached 3703 MWD/T, as logged by VBWR operating personnel. Applying the same scale factor between logged exposure and Ce-Cs analysis of the first fuel sample gives a corrected exposure of 4416 MWD/T. Further debugging of EPITHERMOS, the epithermal extension of the BNL THERMOS code, is in progress. A flux wire exposure is being prepared to map the thermal neutron spectrum in the neighborhood of the test pins in the program fuel element.
Date: October 15, 1963
Creator: Robkin, M. A.
Partner: UNT Libraries Government Documents Department

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 of large "hot spots" continue to appear. The largest spot so far observed was 44 mils long, 20 mils wide, and of the order of 20 mils thick. This spot has a PuO2 concentration which varied from 70% on the periphery to 100% at then center. There is some evidence that the segregated regions are elongated with their long axes perpendicular to the direction of the pressing of the green pellet. Determination of the size and concentration distribution is continuing. The EPITHERMOS code now seems to be operating correctly. A test problem for a typical water lattice converged in eleven iterations. The computation of the spectrum for a pure water medium gave results which agreed very well with the expected 1/E spectrum. At the end of the quarter, the program fuel element had received a cumulative total of 4449 MWD/T exposure. This total is as logged by VBWR operating personnel. Applying the same scale factor, between logged exposure and Ce-Ca analysis of the first fuel sample, gives a corrected exposure of 5306 MWD/T. Three sets of flux wires were successfully irradiated at three thimble locations in the project fuel element. Counting is in progress and the data will be reduced in the next quarter. The program fuel element was removed from the VBWR during the November shutdown at the end of run 165 after a cumulative exposure of about 5000 MWD/T. Fuel ...
Date: January 15, 1964
Creator: Robkin, M. A.
Partner: UNT Libraries Government Documents Department

In-Core Instrumentation Development Program Quarterly Progress Report June - September 1963

Description: Introduction: The objective of Project Agreement 22 is to determine the feasibility of covering the complete reactor neutron flux start range from 10(3) - 5 x 10(13) nv by using in-core chambers. The counting mode of operating will be used at low neutron fluxes and the root mean square voltage fluctuation mode will be used at high neutron flux levels. Experiments have been run utilizing various ion chambers, gases, gas pressures, voltage, and cables to measure sensitivities and range operating in the counting and RMS voltage modes. Theoretical discussions are presented showing how the RMS voltage is related to individual pulse at both amplifier input and output. Noise is also compared at amplifier output so that the optimum bandwidth can be selected. Spectral shifts with changes in applied voltage causing signal variations have been examined and can be eliminated by appropriate selection of amplifier bandwidth. In the counting mode, all experiments have been conducted with unterminated cable. The chamber has been designed with geometry, gas, and pressure to completely stop fission fragments in the gas and hence maximize the charge generated in the chamber. Cables have been selected to minimize capacity. Various gases, pressures, and voltages have been used to determine that an optimum design has been achieved.
Date: October 1963
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department

Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 5

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: January 1, 1964
Creator: Sorlie, T.
Partner: UNT Libraries Government Documents Department

Transition Boiling Heat Transfer Program; Fourth Quarterly Progress Report, October - December 1963

Description: Summary: Heat transfer tests employing the two-rod test section without film tripping devices have been completed. Representations defining critical heat flux, transition boiling and film boiling behavior at high pressures and over a steam quality range of 25 to 90 percent are shown. Fabrication of a new observational test section was completed and initial test results with high-speed motion pictures were obtained. A test loop instability which was found to affect transition boiling behavior was detected and partially eliminated.
Date: January 1, 1964
Creator: Quinn, E. P.
Partner: UNT Libraries Government Documents Department

In-Core Instrumentation Development Program Quarterly Progress Report September - December 1963

Description: Introduction: The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. The counting mode of operation will be used at low neutron flux levels and the RMS voltage fluctuation mode (Campbell Theorem) will be used at high neutron flux levels. The June-September Progress Report (GEAP-4386) shows how the RMS voltage mode can be used, discusses counting problems with long cable and ways of maximizing signal levels. This report discusses primarily the effect of gamma on counting with in-core ion chambers and the range of neutron flux measurable in the RMS voltage mode. Readers are referred to GEAP-4386 for a summary of all previous progress to attain the objective of PA-22.
Date: January 1964
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department

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 round data for corrosion rates and hydriding rates at 300, 400, and 500 degrees C as a function of Nb, Cr, Fe, and Cu content. Solution of the equations for particular service temperatures yield Zr-Cr alloys at optimum at lower temperatures and Zr-Cu-Fe alloys as optimum at the higher temperatures. The second round test results show that neither Ni nor Be additions to Zr-Cr or Zr-Cu improve the performance over that of the optimum Zr-Cr or Zr-Cu-Fe alloys. For the first round heat treatment used, post-corrosion ductility depends on two factors in addition to alloy composition and hydrogen content: crystallographic texture and intermetallic aging reactions. Alloys with a high original ductility are embrittled less by a given amount of hydrogen than are alloys with low original ductility. From the second round tests, it was found that raising the final alpha annealing temperature from 565 to 788 degrees C gives better over-all corrosion, hydriding performance, and resistance to hydrogen embrittlement for both the Zr-Cr and Zr-Cu alloys tested.
Date: January 1964
Creator: Klepfer, H. H.; Jaech, John L.; Blood, R. E.; Perrine, H. E. & Urata, M. E.
Partner: UNT Libraries Government Documents Department

Sodium Mass Transfer. [Part] XI. 1963 Test Run Reports (January - June)

Description: Technical report describing how corrosion data and exposure effects were obtained by subjecting metallic samples, during programmed test runs to flowing sodium in 6 test loops fabricated with various combinations of three selected materials, Type 316 stainless steel, 2 1/4 Cr-1 Mo alloy steel, and 5 Cr-1/2 Mo-1/2 Ti alloy steel. Information produced by each test run, including operational and metallurgical data and analyses, is presented. Data are shown in tables, graphs, and drawings.
Date: February 1964
Creator: Lockhart, R. W.
Partner: UNT Libraries Government Documents Department

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 plastic than mono-clinic zirconia when tested at temperatures above 200 degrees C. The behavior of anion-deficient oxides indicated that they were more plastic than stoichiometric oxides even though the hardness values were identical at 23 degrees C. The former were free from cracks at the indentions, whereas, stoichiometric oxides exhibited extensive cracking around and between indentions. The behavior of actual, thick (72 microns) oxide films during tensile deformation of oxidized metal samples indicated that considerable plasticity occurs in the oxide at 500 degrees C but that the films are brittle at 23 degrees C. It was concluded that the plasticity of the oxide may be greater than that of the oxygen-contaminated substrate at elevated temperatures and may be the means by which epitaxial strains are minimized.
Date: February 20, 1964
Creator: Douglass, D. L. (David Leslie), 1931-
Partner: UNT Libraries Government Documents Department