Technical report describing the four-rod heat flux experiments that are a part of a continuing program of study of the critical heat flux, or burnout phenomenon in order that water cooled reactors can be designed with a maximum of safety and efficiency. During heat transfer with boiling, there is a particular heat flux, for a given set of flow conditions and geometry, above which the nucleate boiling process begins to break down. This breakdown of the nucleate boiling process is known as burnout, critical heat flux, departure from nucleate boiling (DNB), and boiling crisis. The present method at General Electric of avoiding the critical heat flux conditions in the reactor is to limit the heat flux, for a given set of flow conditions, to a fraction of the critical heat flux at the same conditions in the single-rod test section of Janssen and Kervinen. Because the critical heat flux of a heater rod facing an unheated wall is lower than that of a heater rod facing another heater rod, the critical heat flux conditions of the single-rod test section, will be a conservative estimate of the critical heat flux conditions in a multirod reactor. The main purpose of these experiments …

Technical report describing that the pressure drops along 3/4-inch, 1-inch, and 1-1/4 inch straight pipes and across three contraction-expansion inserts in a 1-inch pipe have been measured under both single- and two-phase flow conditions. Pressure was varied from 600 to 1400 psia, flow from 0.25 x 10(6) to 1.66 x 10(6) lb/hr ft, and quality from zero to 90 percent. The single-phase pipe friction factor agrees with the Moody value for smooth pipe. The two-phase friction for horizontal flow shows no size effect in the range of pipe sizes from 3/4 inch to 1-1/4 inch. The values lie below the Martinelli curve at the lower qualities (x<0.6), but at high qualities tend to be above the Martinelli curve. The single-phase loss coefficient for the three contraction-expansion inserts show very little Reynolds number effect in the range of channel Reynolds numbers from 3 x 10(4) to 5 x 10(5). The two-phase data for insert number 1 has not yet been reduced. The two-phase loss for insert numbers 2 and 3 lies generally below the loss prediction based on a homogeneous flow model. The two-phase loss for insert number 2 shows excellent agreement with the corresponding loss for the S-1 insert in …

Abstract: (1) Grain growth in UO2 pellets was studies between 100 C and 2600 C. The pellets were encapsulated in small vacuum-tight tungsten containers in an argon atmosphere. (2) The grain size-time relationship could be expressed by an equation. A low exponent, m>_ 1/3, was found in those experiments and is related to the type of UO2 investigated. An activation energy of 65 kcal/mole was obtained for the grain growth process. The time exponent, m, increased with increasing temperature if the pellets were not contained in closed capsules bu heated under an argon pressure of 1.5 atm. (3) An interaction between tungsten and UO2 could be observed at a a temperature of 2600 C after prolonged heat treatment.

From introduction: "This is the sixteenth 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."

Introduction: The purpose of this technical report is to review the water chemistry methods and equipment developed for use with the Maritime Loop Irradiation Program conducted in the General Electric Test Reactor (GETR) from December 2, 1960 to July 19, 1962. Special emphasis is given to areas having general application to other high purity water systems. The Appendix includes a discussion of specific conductivity and pH in high purity water systems. A major section of this report is devoted to a review of gross activity levels on coupons of two different surface finishes exposed in the loop coolant system for various time intervals. A major objective of the chemistry program was to select or develop analytical methods such that the analyses could be performed at the loop location by technical personnel who normally operate the loop. By this means, frequent samples were obtained and analyzed directly thus providing close monitoring and control of the loop water chemistry at minimum expense.

Summary: The Mixed Spectrum Superheater (MSSR) is an integral superheater reactor in which boiling occurs in an annular Boiling Water Reactor section and steam in superheated in an unmoderated fast section in the center. A Mixed Spectrum Critical Assembly (MSCA) to be operated at the Vallecitos Atomic Laboratory has been designed to mock up a 75-150 MWe prototype MSSR. The principal experimental measurements aimed at proving the feasibility of the MSSR concept include power distribution, Doppler effect, flooding effects, distribution of reactivity, control rod worths, and the effect of the control system on the power distribution.

Summary: Technical report describing the testing of eight Potter Meters, for metering inlet flow and measuring exit steam qualities in the Consumers Big Rock Point Instrumented Fuel Assemblies, were individually calibrated for flow and pressure drop up to 500 gpm in the low temperature (130 F) fluid flow facility. The flow calibration comparison made with an ASME orifice installation, agreed to within + - 1 percent among seven of the meters, and meter Serial No. 8 was 2.8 percent lower than the others. Pressure drop among the meters was within about 5 percent. Locked rotor pressure drop data was obtained on one meter. A fully instrumented fuel bundle was tested in the low temperature facility and pressure drop data obtained for the tieplates and meters, spacers, and channel rods. A mock-up of the exit end of the instrumented fuel bundles, composed of 1 foot of fuel rods, tieplate, and Potter Meter was tested in the High Pressure Heat Transfer Facility. Data was obtained for single- and two-phase calibration of total flow and exit steam quality in an instrumented bundle. Each meter was operated, for a minimum of 6-8 hours after bearing modifications necessitated by seizure of the rotors, in the …

Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. The development of pulsed neutron source techniques for large power reactors has led to a new theoretical model recently developed by E. Garelis and J.L. Russell, Jr. The theory is presently based on a bare, one-group model with m-delayed precursors and takes all spatial modes into account. Results indicate, however, that the application of this model is much broader. Experiments were designed and carried out to both verify this new theory and to demonstrate the performance of the experimental hardware in a large power reactor.

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.

A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. All program efforts have been temporarily deferred except for those associated with the irradiation of the program fuel element in the VBWR. The program fuel element was exposed to a burnup of 831 MWD/T during the quarter which brings the total to 3165 MWD/T. Applying the same scale factor between logged exposure and Ce-Cs analysis of the first fuel sample gives a corrected exposure of 3774 MWD/T.

Abstract: Residual and fission gas release from UO2 were studied in the laboratory and in in-reactor experiments. Arc-fused powder and sintered pellets were used to determine the rate of evolution and types of residual gases as a function of temperature. Fission gas release was related to the average UO2 temperature and fission gas release calculations were made using the latest thermal conductivity values, isotopic half lives, and branching ratios available in the literature. The results obtained were compared with those available in the literature, and a satisfactory agreement was found among the groups of comparable data.

Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities

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 following conclusions are based on the out-of-pile general corrosion and localized attack studies completed to-date on several 300 series stainless steels: (1) Utilizing a sodium chloride-cycle test that produces a type failure that can occur in a superheat reactor system, Types 347 and vacuum-melted 304 SS have failed while vacuum-melted 310 SS was acceptable. (2) An improved chloride cycle test utilizing ferric chloride as the additive has been developed that produces an intergranular type failure similar to that experienced in the fuel cladding failures in the SADE and ESADE facilities. types 304 and 315 SS have failed in the test. (3) Present methods of ultrasonic testing will find through cracks but are not completely dependable for assessing lesser degrees of intergranular attack. (4) It is hypothesized that a definite interplay exists between chemical attack and stress. The application of stress will orient intergranular attack preferentially in a direction perpendicular to the stress.

From introduction: "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."

Abstract: This technical report covers the development work conducted during a planned program with the U.s. Atomic Energy Commission, Contract AT(04-3-189, Project Agreement 22, directed toward the development of high temperature, nuclear radiation resistant, telemetering devices. The development program is devoted to: (1) investigation and selection of two possible telemetering devices, and electromechanical commutating switch and an AM oscillator employing TIMM circuit elements, (2) procuring the electromechanical commutating switch to specification, (3) building and operating a TIMM oscillator, and (4) temperature testing of both devices. A resistance-coupled Wien-bridge sine wave TIMM oscillator was build and tested both as an oscillator, and in combination with other oscillators to simulate a telemetering system. An electromechanical commutating switch rated for 350 F operation, instead of 700 F as originally specified, was procured and tested. The drive motor and gear reduction unit which is designed to drive the commutating switch, is rated for 750 F operation and designed to operate in an nuclear reactor radiation environment of 1 x 10(17) nvt and 1 x 10(10) R.

Quarterly report discussing progress on the Fast Ceramic Reactor Development Program, "an integrated analytical and experimental program directed toward the development of fast reactors employing ceramic fuels, with particular attention to mixed plutonium-uranium oxide" (p. 1).

A program is in progress for the design of a zirconium base alloy for steam service as nuclear fuel cladding. Thirty-one alloys selected for study of corrosion rate, hydriding rate and hydrogen embrittlement are in test. The corrosion testing of 1800 coupons to 3000 hours at at 300, 400, and 500 degrees C in refreshed steam has been completed. Statistical data analysis of the corrosion results are reported and alloys showing better corrosion performance at all test temperatures than that for Zircaloy-1 are discussed. Preliminary data for hydrogen uptake after long exposures at 400 and 500 degrees C are presented; the uptake for alloys showing the best corrosion performance is discussed. Post-corrosion mechanical property measurements are also reported along with the preliminary results of x-ray diffraction and metallographic studies relating to hydrogen embrittlement. A wide variation in resistance to embrittlement at a given hydrogen level was observed and can be tentatively correlated with original ductility, crystallographic texture, and hydride platelet orientation. The testing of a second round of ten alloys is also in progress. Studies concerning the mechanism of corrosion and hydriding in zirconium alloy are also reported. The results of recent neutron activation analyses of stripped corrosion films are …

Introduction: The Transition Boiling Heat Transfer Program is sponsored jointly by the USAEC and Euroatom and is being conducted by the General Electric Company. The work commenced on this program February 11, 1963. The objective of this program is to perform basic investigation and measurement of the transition boiling regime in high pressure bulk boiling water flows, with particular emphasis i the high range of steam qualities.

Task A: Modification and Preparation of Experimental Facility. Facility engineering and layout is about seventy-five percent complete. Task B: Design and Construction of Test Sections. The major dimensions and characteristics of the metal and glass test sections have been calculated. One feasibility test of the electrically conducting coating on samples of glass tubing has been completed. Task C: Design and Construction of Test Stand, Task E: Pressure and Temperature Instrumentation for Test Section and Task F: Power Supply for Test Section. Preliminary engineering has been initiated on these tasks. The planned approach has been defined in each case. For Task E the transducer specifications have been defined and quotations on and/or sample units of the transducers have been requested. Tasks C and F can proceed with detailing as soon as drafting on Task B is about 50 percent complete. This point is scheduled to be reached during the first part of July. Task D: Void Fraction Instrumentation. The requirements for the x-ray instrumentation have been considered in the course of Task B and the x-ray power supply is presently on hand. The detailed engineering effort on this task is not scheduled to begin before July.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

A potential performance limitation of superheat fuel is the susceptibility of the fuel cladding to low cycle fatigue failure. Two simplified analytical methods are presented to estimate the cyclic lifetime of circular superheat fuel cladding. One failure relation is based on a displacement method. The other failure relation is based on a stress method. These relations were compared with data from the literature, and with data involving damage obtained by Reynolds. A recommended design procedure involving the relations is presented. The technique was applied to the SADE 4B experiment with moderate success. These cycling relations involve only mechanical damage imposed by cycling, with a modification for additional damage caused by radiation; they do not include any other potential performance limiting mechanisms, such as stress corrosion, which are normally factored into the over-all fuel design. This work work done under Task C (Materials Development) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189 - Project Agreement 13.

Tests were run at the General Electric Company, Atomic Power Equipment Department, to determine the burnout conditions for a non-uniformly heated rod in an annular geometry.

A process was developed in which stainless steel-clad UO2 fuel rods are fabricated by high-temperature coextrusion. The process has a potential of being a more economical method for the preparation of stainless steel-clad UO2 fuel rods than the conventional pellet process. Consequently, it was considered advantageous to evaluate the irradiation characteristics of fuel rods fabricated in this manner. Therefore, 24 coextruded fuel rods were manufactured for evaluation in a reactor. The required amounts of UO2 and clad were soaked in separate containers at 1875 and 760 degree C, respectively. The containers were removed from their respective furnaces and were coextruded in one pass. A force of 450 to 475 tons was used, and a reduction ratio of 18 to 1 was obtained. The coextruded rods were cut to the approximate length, and the ends were sealed with an acid-resistant tape. The carbon steel can covering the stainless steel clad was removed by immersion in 1:1 nitric acid for 20 minutes. The rods were visually inspected, the specified lengths of clad and fuel were obtained by machining, and the correct diameter was obtained by belt sanding. The fabrication of the fuel rods was completed by inserting the plenum support tubes and …