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.

Abstract: The physical property variation of superheated steam differs sufficiently from most other gases to warrant experimental investigation of heat transfer performance. Results are reported here of measurements made in a uniformly heated circular duct with steam at 1000 psi. The data agree very well with the expression use for design purposes, which is based on information in the literature for heating of other gases as well as steam. This work was a continuation of that performed under Task (Heat Transfer) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189, Project Agreement 13.

This technical report describes the development, design, operation, and performance of a full-circle, radial-vane steam separator for the boiling water section of a nuclear superheat reactor. Steam-water tests of this model have demonstrated that is has vane capacity in excess of that required for the 300-Mx(e) separate superheat reactor and for the 300-Mw mixed spectrum superheat reactor. It is proposed that the vane capacity requirement of the 600 Mw(e) separate superheat reactor may be attained by increasing the nozzle length.

Technical report describing that void measurements were made in the 1/2-inch by 1-3/4-inch rectangular channel, for both flow up and flow down, at pressures of 600, 1000, and 1400 psia, and at various flows and quantities. Results at 1000 psia and 20 percent quality show that for the lowest flow both the void distribution and the average void are much different for flow down than for flow up, the void fraction for flow down being much higher. However, when the flow is increased both the void distribution and average void for flow down tend to approach the corresponding values for flow up. At 1000 psia, both flow up and flow down, the void fraction for 5 percent quality increases gradually from the wall to the center of the channel, and peaks at the center. At 20 percent quality, the void fraction increases abruptly from the wall and tends to be constant over the middle 65 percent of the channel. the void fraction for flow down is always greater than for flow up, other things being equal.

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

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 …

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 …

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.

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: A prototype control rod containing absorber plates made from an electro- deposited dispersion of boron carbide in nickel was tested in the VBWR. It was exposed to the reactor environment of 545 degree F boiling water and thermal neutron fluxes (perturbed) which ranged from 0.6 to 1.1 x 10/sup 13/ nv for 2236 hours over a period of six months. The maximum B/sup 10/ burnup achieved during the test period was 1.8 percent. After irradiation, the rod was examined. The results of the examination are summarized below: (1) The B/sub 4/C-- Ni plate assembly did not undergo significant dimensional changes during irradiation. (2) Numerous blisters developed on both the outer and inner surfaces of three of the four plates. Blistering was more severe on the outer surface than on the inner, and was most severe in a large region located in the lower half of plate 4. Metallographic examination revealed that the blisters were located only in the 2- mil protective nickel overlay covering the B/sub 4/C-- Ni dispersion. It was concluded that they formed from the buildup of gas pressure at the Ni: Ni-- B/sub 4/C interfaces, rather than from corrosion attack. Helium from the B/sup 10/(n alpha …

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).

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

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.

Introduction: The object of Project Agreement 22, Task 1, is to develop improved detectors which can operate up to 1000 F for in-core power monitoring. Several ideas have been developed to achieve this goal: (1) root mean square fluctuation voltage measurement of ion chamber signals, (2) thermocouple-type detectors, and (3) fabrication developments.

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.

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.

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.

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."

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.

The General Electric Company is designing and building a 180,000 kilowatt nuclear power plant for the Commonwealth Edison Company at a site near the confluence of the Kankakee and Des Plaines Rivers in Grundy County, Illinois, about 47 miles southwest of Chicago. The plant will be known as the Dresden Nuclear Power Station, and will employ a nuclear reactor of the dual-cycle boiling water type.

This report is submitted to the U. S. Atomic Energy Commission as a final summary safeguards and hazards evaluation of a proposed test reactor at its Vallecitos Atomic Laboratory in Alameda County of California. It is the purpose of this report to provide sufficient data to obtain an AEC facility license for the reactor.

In concurrent flow of two phase mixtures there exists a velocity difference between the vapor and liquid phases. This difference in velocity is known as the slip velocity. The prediction of slip is the subject of Part I. In boiling systems there is some heat transfer rate at which nucleste boiling becomes unstable. At this point the separate bubbles coalesce forming an insulating vapor film on the heat transfer surface resulting in the destruction, or burnout, of the heater. The prediction of the conditions causing burnout is the subject of Part II.

The General Electric Company is designing and building a 180,000 kilowatt nuclear power plant for the Commonwealth Edison Company at a site 47 miles southwest of Chicago. The construction permit was issued on May 4, 1956 but is subject to submittal to the Commission of a final hazards summary report and a finding by the Commission that the final design provides reasonable assurances that the health and safety of the public will not be endangered by operation of the reactor in accordance with the specified procedures.