Design and Operating Experience of the ESADA Vallecitos Experimental Superheat Reactor (EVESR) (open access)

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."
Date: February 1965
Creator: Ianni, P. W.; Fritz, J. R. & Law, D. D.
System: The UNT Digital Library
Pressure Vessel Steel Surveillance Program for General Electric Power Reactors (open access)

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."
Date: January 8, 1965
Creator: Brandt, F. A.
System: The UNT Digital Library
Fuel Cycle Program Progress Report: Fifteenth Report, January-June 1964 (open access)

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.
Date: July 15, 1964
Creator: Howard, C. L.
System: The UNT Digital Library
Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: April 1 - June 30, 1964 (open access)

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.
Date: July 15, 1964
Creator: Robkin, M. A.
System: The UNT Digital Library
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 7 (open access)

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.
Date: July 1, 1964
Creator: Howard, C. L.
System: The UNT Digital Library
Southwest Experimental Fast Oxide Reactor Development Program: First Quarterly Report, April-June 1964 (open access)

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."
Date: July 1964
Creator: Horst, K. M.
System: The UNT Digital Library
Transition Boiling Heat Transfer Program; Sixth Quarterly Progress Report, April - June 1964 (open access)

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.
Date: July 1, 1964
Creator: Quinn, E. P.
System: The UNT Digital Library
UO2 Pellet Thermal Conductivity From Irradiations With Central Melting (open access)

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 …
Date: July 1964
Creator: Lyons, M. F.; Coplin, D. H.; Pashos, T. J. & Weidenbaum, B.
System: The UNT Digital Library
Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Seventh and Eighth Quarterly Progress Report, October 1, 1963-March 31, 1964 (open access)

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.
Date: April 24, 1964
Creator: Garelis, Edward & Meyer, P.
System: The UNT Digital Library
High Performance UO2 Program Quarterly Progress Report No. 12 January-March 1964 (open access)

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.
Date: April 15, 1964
Creator: Weidenbaum, B.
System: The UNT Digital Library
Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: January 1 - March 31, 1964 (open access)

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.
Date: April 15, 1964
Creator: Robkin, M. A.
System: The UNT Digital Library
Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 6 (open access)

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.
Date: April 1, 1964
Creator: Howard, C. L.
System: The UNT Digital Library
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 (open access)

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.
Date: April 1964
Creator: Zuber, N. & Findlay, J. A.
System: The UNT Digital Library
Fabrication of fuel Cladding From Incoloy Alloy 800 : an Evaluation of Methods (open access)

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 …
Date: April 1964
Creator: Kirby, R. F.; MacMillan, D. F. & Punches, J. R.
System: The UNT Digital Library
High Power Density Development Project: Sixteenth Quarterly Progress Report, January-March 1964 (open access)

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.
Date: April 1, 1964
Creator: Holladay, R. L.
System: The UNT Digital Library
In-Core Instrumentation Development Program Quarterly Progress Report January - March 1964 (open access)

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.
Date: April 1964
Creator: DuBridge, R. A.
System: The UNT Digital Library
Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Tenth Quarterly Report, January-March 1964 (open access)

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.
Date: April 1964
Creator: Breizy, C. E.
System: The UNT Digital Library
Transition Boiling Heat Transfer Program; Fifth Quarterly Progress Report, January - March 1964 (open access)

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
Date: April 1, 1964
Creator: Quinn, E. P.
System: The UNT Digital Library
A Program of Two-Phase Flow Investigation Quarterly Report: Fourth Quarterly Report, January-March, 1964 (open access)

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 …
Date: March 16, 1964
Creator: Staub, F. W. & Zuber, N.
System: The UNT Digital Library
Accurate Nuclear Fuel Burnup Analyses; Ninth Quarterly Progress Report, (December 1963 - February 1964) (open access)

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.
Date: March 1, 1964
Creator: Rider, B. F.; Peterson, J. P., Jr.; Ruiz, C. P. & Smith, F. R.
System: The UNT Digital Library
AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel (open access)

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 …
Date: March 1964
Creator: Ogawa, S. Y.
System: The UNT Digital Library
EVESR Nuclear Superheat Fuel Development Project: Seventh Quarterly Report, December 1963 - February 1964 (open access)

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.
Date: March 1964
Creator: Pennington, R. T.
System: The UNT Digital Library
General Corrosion of Incoloy-800 in Simulated Superheat Reactor Environment (open access)

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. …
Date: March 1964
Creator: Pearl, W. L.; Brush, E. G.; Gaul, G. G. & Wozadlo, G. P.
System: The UNT Digital Library
Reactor Pressure Vessel Material Surveillance Program at the Garigliano Nuclear Power Plant (open access)

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.
Date: March 1964
Creator: Brandt, F. A. & Kobsa, I. R.
System: The UNT Digital Library