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Subcooled burnout at high flows
Task II-B of the High Performance UO2 Program calls for testing a series of UO2 fuel rods in the PW loop of the GETR. The purpose of the tests is to obtain a relationship between the extent of central melting and the length of time before the fuel rod fails as a result of interaction between the fuel and the cladding.
Design and Operating Experience of the ESADA Vallecitos Experimental Superheat Reactor (EVESR)
Summary: "The various design features significant to superheat are described for the 12-1/2 MW., 960 psig, 1050° F, steam-cooled, low-enriched, annular-fueled, experimental superheat reactor built by the General Electric Company at the Vallecitos Atomic Power Laboratory. Results obtained during the first six months of full-power operation, on the emergency cooling system, core thermal performance, and pressure vessel temperatures are presented and compared with predictions. Operating experience with over-all reactor system is also discussed."
Pressure Vessel Steel Surveillance Program for General Electric Power Reactors
Abstract: "Pressure vessel steel surveillance programs are performed in nuclear power reactors to provide knowledge of the mechanical properties of the pressure vessel material as neutron irradiation proceeds. A standard surveillance program is described. Design of specimens, capsules, and associated equipment, as well as selection of test material and techniques for special preparation and testing, are discussed."
Fuel Cycle Program Progress Report: Fifteenth Report, January-June 1964
Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities.
Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: April 1 - June 30, 1964
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: Project fuel fins irradiated to 1860, 3000, and 5300 MWD/T have been successfully sampled in the Radioactive Materials Laboratory. The samples have been dissolved and aliquots delivered to Chemistry for Mass Spectrometry and burnup determination. The first Stanford Pool Irradiation indicated that there was some inconsistency in the thermal flux and the near thermal epithermal flux. The experiment was repeated, increasing the number of foil wheel positions from two to three. The data from the second measurement are being reduced. The EPITHERMOS code modification has been completed. Comparisons between the results computed by the code and experimental data show much improved agreement. The metallographic photomicrographs of a polished half-pellet from rod F, irradiated to 5000 MWD/T, show structure very similar to that shown by the pellet from rod S, irradiated to 1860 MWD/T.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 7
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). One task is in progress: Task I - Data Logging and Computer System. The work on the other tasks is being planned and initiated.
Southwest Experimental Fast Oxide Reactor Development Program: First Quarterly Report, April-June 1964
From abstract: "This report summarizes the work performed on the Research and Development Program for the Southwest Experimental Fast Oxide Reactor."
Transition Boiling Heat Transfer Program; Sixth Quarterly Progress Report, April - June 1964
Summary: Transition boiling data was taken with an improved flow loop, to explore the influence of loop characteristics on rod temperature fluctuations the transition region was found to be much smaller than for comparable conditions with a different loop. Also the amplitude, and frequency of the temperature oscillations, were significantly less than before. These results indicate that loop characteristic and flow disturbance parameters play a prominent part in governing the transition temperature fluctuations. Additional two-rod transition boiling data are presented. The results include data taken at high wall temperature levels during a demonstration test at low steam qualities, and the effect of a change in rod spacing on heat transfer performance.
UO2 Pellet Thermal Conductivity From Irradiations With Central Melting
Abstract: Continued irradiation experience under the AEC - Euratom, UO2 High Performance Program provided five separate and distinct sets of data on UO2 thermal conductivity. Four of these results are expressed in terms of the value of the thermal conductivity. The first two of these measurements were applicable -- strictly -- to poly crystalline UO2. Recently, three additional sets of measurements have been obtained -- all pertinent to UO2 after the formation of large columnar grains. The extent of melting in the experiments on which the results are based ranges from slight, to greater than 70 percent of the fuel cross section. The conclusions from all of these thermal conductivity measurements considered together are: (1) The true value of the UO2 conductivity integral form 0 degrees C to melting (2805 - 15 degrees C) lies in the range from 90 to 96 W/cm. The most probable value is closer to 90 W/cm. To ensure no central melting and the associated clad swelling the maximum thermal performance level for solid pellet, UO2 fuel rods should not exceed 90 W/cm. (2) Any improvement in thermal conductivity due to the formation of large, columnar UO2 grains is small and not detectable within the experimental accuracy of the measurement, i.e., 3 to 4 W/cm.
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Seventh and Eighth Quarterly Progress Report, October 1, 1963-March 31, 1964
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. In the course of this program, a new theory was suggested and an experimental apparatus was designed and built. Experiments were carried out to test the new model. This present report contains additional data and information extracted from the experiments at PG&E Humboldt Bay Power Reactor at Eureka, California. During the last days of 1963 a number of control rod and fuel bundle worth measurements were made in the ESADA Vallecitos Experimental Superheat Reactor (EVESR) using the (k[beta]/[script l] technique. A description of the experiments is given in the text of the report and some results are reported. A computer program was written to perform the data analysis of the pulsed neutron experiments and the code is discussed in the Appendix.
High Performance UO2 Program Quarterly Progress Report No. 12 January-March 1964
Work performed during the quarter is summarized by: direct measurement of fission gas pressure, loop operations, performance of UO2 fuel, UO2 grain growth and melting studies.
Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: January 1 - March 31, 1964
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: Work has begun in the Radioactive Materials Laboratory to sample the project fuel from the pins irradiated to 1800 and 5000 MWT/T. Some delay has been experienced due to preemption of the hot cells by priority work. Examination of the autoradiographs of the un-irradiated project fuel showed that in a volume of fuel approximately equivalent to a pellet there were 13 hot spots larger than 15 mils. Evaluation of these spots with the fuel analyzer showed that they contained about 14 mg of PuO2 or about 9% of the total present. The EPITHERMOS code is being modified to automatically normalize the epithermal scattering to the correct value for all moderators. Calibration of the flux wires has been made and the reduction of the data from the VBWR irradiation is nearly complete. A similar resonance activation was made in the water reflector of the Stanford Pool Reactor to obtain the relative activity in a well-defined pure water spectrum. Reduction of these data is also in progress.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 6
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.
The Effects of Non-Uniform Flow and Concentration Distributions and the Effect of the Local Relative Velocity on the Average Volumetric Concentration in Two-Phase Flow
Abstract: A general expression which can be used either for predicting the average volumetric concentration or for analyzing and interpreting experimental data is derived. The analysis takes into account both the effect of non-uniform flow and concentration profiles as well as the effect of the local relative velocity between phases. The first effect is taken into account by a distribution parameter, whereas the latter is accounted for by the weighted average drift velocity.
Fabrication of fuel Cladding From Incoloy Alloy 800 : an Evaluation of Methods
Summary: On the basis of its high temperature, physical and corrosion properties, Incoloy Alloy 800 was selected as a candidate for fuel cladding nuclear superheat applications. At the time of its selection, there was little information or experience with Incoloy 800 in the production of thin-walled, small diameter tubing suitable for nuclear fuel cladding. As a result, special purchasing efforts were required for the procurement of initial tubing used in fuel fabrication. As-received welded and drawn tubing proved to be generally good but showed some conditions which were undesirable, the major one being lack of complete recrystallization and homogenization of the weld zone. The possible effect of this condition upon the fuel performance was not immediately known; however, subsequent development work indicated that the non-homogeneity of the weld could affect adversely its mechanical and corrosion properties in relation to the parent metal. A development program was initiated to determine treatment sequences suitable for the fabrication of welded and drawn tubing with a fully recrystallized and homogenized weld structure. This was accomplished by butt welding lengths of Incoloy strip which were subsequently cold rolled and annealed to simulate tube fabrication steps. Requirements imposed on this work were that all processes developed must be amenable to normal production equipment and procedures used in commercial tube manufacturing. As a result of the experiments undertaken, a sequence of cold drawing and annealing steps was established suitable for recrystallization and homogenization of weld zones in welded-and-drawn tubing. In order to obtain complete sequential chemical and metallurgical history of nuclear grade Incoloy 800 tubing, a 3000-pound ingot was purchased to the required chemical specifications and reduced to tubing through both seamless and welded drawn routes. The management of the sequential metallurgical steps needed for tubing fabrication resulted in establishing a workable process through a series of …
High Power Density Development Project: Sixteenth Quarterly Progress Report, January-March 1964
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, (2) Task 1B-Fuel Fabrication Development. Assembly, (3) Task II-Stability, Heat Transfer and Fluid Flow, (4) Task III-Physics Development, and (5) Task IV-Co-Ordination and Test Planning.
In-Core Instrumentation Development Program Quarterly Progress Report January - March 1964
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. This technical report discusses the following topics: low versus high cable termination impedance, amplifier considerations, noise considerations, gas and pressure selection, cable selection, effect of gamma, effect of temperature, and remaining problems.
Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Tenth Quarterly Report, January-March 1964
Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, transient testing of fuel, fuel performance evaluation, fast-flux irradiation of fuel, and reactor physics and core analysis.
Transition Boiling Heat Transfer Program; Fifth Quarterly Progress Report, January - March 1964
Summary: Tests with the two-rod assembly were performed with liquid film trippers attached to the unheated wall, and a variation in rod spacing. Experimental data and improved high-speed motion pictures have been obtained of transition boiling behavior. The changes of the local heat transfer process between nucleate and film boiling can be readily distinguished i the motion pictures. Observational test performed with very short fins on the heated surface resulted in essentially eliminating transition boiling temperature fluctuations and doubling the film boiling coefficient. These gains were attained without reduction of the critical heat flux
A Program of Two-Phase Flow Investigation Quarterly Report: Fourth Quarterly Report, January-March, 1964
Summary: The design, construction and assembly of all components were completed during the first contract year previous to December 1963. These efforts, defined by Tasks A-F, are document in (1), (2), and (3). Brief summaries of these completed efforts are given in the introduction to each of the tasks in the text of this report. The digest given below covers only the shakedown and analysis work carried out in the fourth quarter of the first contract year. Task G Shakedown Tests. The photographic procedure has been experimentally defined for the glass test section. Four automatic 35 mm cameras and four strobe light sources have been ordered on ATL funds and their respective mounting arrangements are in place. Roughly ten test runs were carried out in the glass test section during the course of the above work. Satisfactory recorder traces have been obtained on all measurement systems. These systems presently meet the accuracy and linearity specifications initially set. An x-ray void fraction signal adjustment and filtering circuit has been design and installed to provide equal resolution across the test section. Calibration disc inserts have been installed to permit satisfactory beam intensity calibration. Good agreement has been obtained between calculated and measured beam intensity profiles across the test section. To improve the flexibility of the differential pressure measurement system, zero suppression up to 1.5 psi has been calibrated for each bridge circuit. This makes up for the fact that the reference pressure control regulator does not quite meet the vendors dead band specifications. To facilitate the initial records chart calibration for the temperature sensors a special precision decade box switching circuit has been installed. While no significant temperature sensor calibration shift has taken place in two months, the lack of adequate reliability of these units will require a periodic calibration check. Negotiation …
Accurate Nuclear Fuel Burnup Analyses; Ninth Quarterly Progress Report, (December 1963 - February 1964)
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.
AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel
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.
EVESR Nuclear Superheat Fuel Development Project: Seventh Quarterly Report, December 1963 - February 1964
Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.
General Corrosion of Incoloy-800 in Simulated Superheat Reactor Environment
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 exposed in the Vallecitos boiling water reactor superheated steam loop indicated the questionable dependability of such stainless steels for this SHX fuel cladding application. The following conclusions are based on the out-of-pile general corrosion evaluations completed to date on Incoloy-800 as a fuel cladding for nuclear superheat applications: 1. The corrosion data from 4000-hour heat transfer tests indicate good corrosion resistance up to at least a 1300°F metal temperature. By use of a devised method of data treatment, the general corrosion for three years exposure at 1300°F can be calculated to average 0.0016 inch with an upper 95 percent confidence limit of 0.0033 inch. 2. A compositionally-disturbed layer develops at the metal-oxide interface. The disturbed layer depth is a function of time and temperature of exposure. The most pessimistic interpretation of the growth rate with time would indicate a layer depth of 0.0017 inch after a three years' exposure at 1300°F. The possible effects of such a layer on cyclic type exposures (i.e., cyclic strain, corrosion fatigue, etc.) have not been evaluated. 3. The corrosion results indicate excellent corrosion resistance of Incoloy-800 when exposed isothermally to steam generated from a simulated BWR and superheated to 1150°F. Corrosion rates .of 3.6 +- 1.2 mg/dm(2) -mo at 1050°F and 10 +- 18 mg/dm(2)-mo at 1150°F are applicable in such exposures with a steam velocity of 20 ft/sec. 4. …
Reactor Pressure Vessel Material Surveillance Program at the Garigliano Nuclear Power Plant
Abstract: A materials exposure program has been established in the Garigliano Nuclear Power Plant to measure the effect of neutron irradiation and time at temperate on the mechanical properties of the reactor pressure vessel steel. Base metal specimens were made from portions of the pressure vessel steel, and weld heat-affected zone and weld metal samples were taken from a weldment made from the pressure vessel steel and simulating a pressure vessel circumferential weld since there are no longitudinal welds in the forged ring shell. The specimens were sealed in helium-filled capsules and placed in the reactor vessel at locations where they will be exposed to a variety of conditions. Tensile property changes will be measured by pre- and post-irradiation tests on small tensile specimens. Fracture characteristic changes will be measured in similar fashion by Charpy V-notch impact tests. The program is planned to cover a 32-year period, with specimens to be removed for test at intervals of 1, 2, 4, 8, 16, and 32 years.
Transition and Film Boiling Data at 600, 1000, and 1400 PSIA in Forced Convection Heat Transfer to Water
Summary: Data were obtained in a two-road test section which consisted of two 7/16-inch diameter heater rods inside a roughly rectangular flow area. The heated length of the rods was 30 inches, with a 15-inch unheated calming length preceding it. Heater wall temperatures were recorded while the heater tubes were trans-versing the critical heat flux and transition boiling; these temperatures were used to calculate heat transfer coefficients. The following general results were obtaining: (a) Pressure has very little effect on the heat transfer coefficient in transition an film boiling. (b) Heat transfer coefficients during film boiling increase with mass velocity and steam quality. (c) The range of film boiling convective heat transfer coefficients observed was 364 to 1150 Btu/h-ft(2)-degrees F. (d) Temperature oscillations occur during transition boiling with a magnitude of as much 700 degrees F, at a frequency of about 1/2 cps. These temperature oscillations are reduced in magnitude as the steam quality and mass velocity are increased, becoming small (~20 degrees F) at high qualities and mass velocity. (e) A preliminary correlation of heat transfer coefficient data correlates the experimental data within about 20 percent. (f) Temperatures rises during transition boiling can be described analytically.
Two-Phase Pressure Losses Quarterly Progress Report: Eighth Quarter, November 12, 1963 - February 11, 1964
Technical report describing that voids were measured in a ½-inch by 1-3/4-inch channel with the S-1 insert (B(0)/B(1) = 0.4, L(0) = 0.1 inch), at 2 inches ahead of the insert (position A), ½-inch past the insert (position B), 5 inches past (position C), and 12 inches past (position D). The conditions were: P – 1000 psia, G = 1.00 x 10(6) lb/h-ft(2), and x = 18.8 percent. Average void and void distribution at position A are the same as for flow in a straight channel. Void distribution at position B shows that the stagnation region downstream of the inserts contains a high fraction of voids. Average void and void distribution at positions C and d show that the two-phase mixture becomes strongly mixed (homogenized) as a result of passing through he contraction-expansion inserts. Distribution at position D approaches the distribution at position A; i.e., the straight channel distribution.
Oxidation Mechanism of Zirconium and Its Alloys. [Part] II. Oxide Plasticity
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.
Nuclear Superheat Quarterly Project Report: Eighteenth Quarter, November, 1963-January, 1964
From introduction: "This is the eighteenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."
Design and Fabrication of Fuel Rods Containing Sintered UO2 Extrusions - Assembly 11L
The extrusion forming of ceramic powders may be economically interesting in the field of nuclear fuel fabrication. When applied to the forming of rod-type uranium dioxide fuel, extrusion processes have been able to produce cylindrical bodies with length-to-diameter ratios much greater than those of the conventional die-pressed pellets. Furthermore, after being sintered, the extrusions have exhibited densities at least as high as those of sintered pellets. Thus, extrusion forming may offer reductions in handling during fabrication and, at the same time, provide a fuel with improved performance characteristics by decreasing the number of discontinuities in the fuel column. This report reviews the production of these extrusions, sets forth some of their characteristics, describes the materials and processes employed in cladding them, and records the pre-irradiation data pertaining to the finished fuel rods and fuel assembly. Irradiation of the fuel assembly in the VBWR was initiated on July 17, 1962.
Design and Fabrication of Pellet Fuel Rods Clad With Thin Wall Stainless Steel
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.
Localized Corrosion of Stainless Steels and High-Nickel Alloys in Simulated Superheat Reactor Environment
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 boiling water reactor quantities of oxygen and hydrogen. During the test, the heater sheaths were exposed through several cycles to saturated steam (with its accompanying moisture carryover) and superheated steam. Failure of Type-304 stainless steel was obtained in periods of less than two weeks; the failures were predominantly transgranular. Type-347 and vacuum-melted Type-304 stainless steels failed in this NaCl-cycle test while Inconel-600, Incoloy-800, Hastelloy-X, Type-406 stainless steel, and vacuum-melted Type-310 stainless steel were acceptable. An improved chloride cycle test with 0.5 ppm chloride added as ferric chloride to the recirculating water was developed. An intergranular failure was obtained similar to that experienced in the superheat fuel cladding failures in the superheat in-pile loops in the Vallecitos Boiling-Water Reactor. Sensitized Type-304 and Type-316 stainless steels failed intergranularly in this test. Inconel-600, Incoloy-800, and vacuum-melted Type-310 stainless steel did not fail when exposed to the test for much longer time periods. During the development and performance of the cycle runs, the superheat facilities were exposed to a myriad of conditions within the extremes of the test parameters involved. Intergranular chemical attack was experienced essentially independent of stress, but the attack was generally distributed. In the presence of high stress, the intergranular attack was more localized and advanced normal to the stress. It is hypothesized that definite interplay exists between chemical attack and stress, and that the …
Mechanical, Fluid Flow, and Heat Transfer Out-Of-Pile Tests on EVESR MKI Prototype Fuel Bundle
Summary: An EVESR MKI prototype fuel bundle was fully instrumented and operated intermittently for a 5-month period at the Pacific Gas and Electric Company’s Moss Landing Power Station. The vessel was operated up to 1000 psi with steam flows from 3000 to 26,600 lb/h, and steam inlet temperatures up to 825 degrees F. Data was recorded for blowout, vibration, flow distribution, heat transfer and pressure drop. The mechanical integrity of the fuel bundle, riser, and jumper system was satisfactory and considered to be of adequate design. No significant vibrations were noted during the various phases of operation. Average flow distribution in three of the inner tubes showed an average variation of 5 percent from equal distribution. The center and corner tubes were low and the side tube was high. Maximum deviation, from an equal one, measured 12 percent. Blowout of the flooded fuel bundle was accomplished with dry or significantly wet 1000 psia inlet steam, that steadied out to a minimum flow of 1250 lb/h. Blowout times were estimated at less than a minute for all flows above 1250 lb/h, and times in the vicinity of 2000 lb/h were estimated to be in the order of 5 to 15 seconds. Once the bundle was blown out a flow of 700 lb/h was sufficient to keep the fuel passages clear. This was true even with steam estimated at 10 to 20 percent wet. Flows below 1250 lb/h caused partial blowout. Usually the A tube blew out first and the B and C tubes gradually cleared as flow was increased. Los of flow then caused comparatively sudden flooding ranging from less than 2 to 3 seconds to several minutes for the different tubes within the bundle. Once the bundle was blown out and the flow maintained at more than 700 lb/h, not …
Preoperational Power Stability Analysis of the Consumers Big Rock Point Plant
Summary: An analytical study of the stability of the Big Rock Nuclear Reactor has been performed for the plant as built, and supplements a previous design stability study. The plant has been determined by this analysis to be very stable under every mode of operation anticipated during Phase I of the development program testing. Even under conservative assumptions of system parameters the minimum calculated gain and phase margins do not go below 13.0 db and 46 degrees, respectively. (Nor are these both reached simultaneously for the same operating condition.) These are characteristics of a very stable, well-behaved system. In addition to this analysis, a second, less conservative series of computations was performed to provide expected realistic closed loop data for comparison with Phase I test results. The most responsive test thus predicted occurs at 60 percent power, 1500 psia, minimum flow, and maximum subcooling. For this case the closed loop peak response of power to reactivity occurs at a frequency of 0.90 cycles per second with an amplitude of 3.90 db. This corresponds to an expected open loop gain margin of 16.5 db and a phase margin of 63 degrees. Although knowledge of reactor transfer function is to be determined from tests designed around the analyzed cases, no operational cases will approach an unstable or even a substantially underdamped situation, according to the analytical predictions.
Sodium Mass Transfer. [Part] 8. Corrosion of Stainless Steel in Isothermal Regions of a Flowing Sodium System
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.
Sodium Mass Transfer. [Part] XI. 1963 Test Run Reports (January - June)
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.
Fuel Cycle Program Progress Report: Fourteenth Quarter, October-December 1963
Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities.
High Performance UO2 Program Quarterly Progress Report No. 11 October-December 1963
Work performed during the quarter is summarized by: direct measurement of fission gas pressure, loop operations, performance of UO2 fuel, UO2 grain growth and melting studies.
Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: October 1 - December 31, 1963
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 …
Consumers Big Rock Point Nuclear Power Reactor Stability Analysis
This report presents the results of an analysis which was undertaken to investigate the power stability of the Consumers Big Rock Point Nuclear Power Reactor.
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 5
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.
High Performance UO2 Program Quarterly Progress Report No. 13 April-June 1964
Work performed during the quarter is summarized by: loop operations, performance of UO2 fuel, UO2 grain growth and melting studies, fuel rod mechanism failure studies.
High Power Density Development Project: Fifteenth Quarterly Progress Report, October-December 1963
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. All fuel irradiation has been terminated with the final shutdown of the VBWR. The high burnup average achieved by a single assembly in the group is 10,000 MWD/T (assembly 1F). Twenty-one of the original 24 assemblies have failed or are suspected of failure. Profilometer tests rung on HPD assembly 2E, Rod B, indicate that localized clad deformation occurs during operation. (2) Task 1B-Fuel Fabrication Development. Assembly. All fuel irradiation has been terminated with the final shutdown of the VBWR. The highest average burnup achieved by a single assembly in the group was assembly 4S with 8400 MWD/T. All assemblies in the group have failed or are suspected of failure. The Phase I developmental fuel continues to be irradiated in the Big rock Point reactor with the lead assembly having reached 1500 MWD/T. Fifteen phase II developmental assemblies are being construction for insertion at Big Rock Point in March. Engineering is underway to provide one instrumented assembly probe and two spare flowmeters for use in phase II testing. Flowmeter bearing are being redesigned to minimize crud access and changes of bearing seizure. (3) Task II-Stability, Heat Transfer and Fluid Flow. Phase I of the reactor performance tests has now been completed. These tests consisted of core performance, control rod oscillator, pressure transient, and flow tests. Reduction of the data from these tests has begun, and preliminary results have been prepared for use by the Consumers Power Company in relicensing for Phase II. (4) Task III-Physics Development. Power distribution calculations have been performed for the proposed 84-bundle, 75 MWe core and for the high …
In-Core Instrumentation Development Program Quarterly Progress Report September - December 1963
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.
Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Ninth Quarterly Report, October-December 1963
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.
Specific Zirconium Alloy Design Program Quarterly Progress Report: Seventh Quarter, October - December, 1963
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.
Transition Boiling Heat Transfer Program; Fourth Quarterly Progress Report, October - December 1963
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.
Accurate Nuclear Fuel Burnup Analyses; Eighth Quarterly Progress Report, (September - November 1963)
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.
EVESR Nuclear Superheat Fuel Development Project: Sixth Quarterly Report, September - November, 1963
Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.
Influence of the Doppler Effect on the Meltdown Accident
The influence of the Doppler effect in the core disassembly process following a meltdown accident is examined with a Bethe-Tait type model in which the Doppler effect, as well as core disassembly, is considered in the reactor shutdown process. It is shown that a strong negative Doppler effect can radically reduce the explosive energy release in such an accident. (auth)
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