This is a report of work completed during the period October 1, 1961, to December 31, 1961. The Program fuel element was completed and assembled. After extensive analysis, it was concluded that the fuel was satisfactory for use in the Program.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the second quarter of fiscal year 1962. The data are summarized in Section II. Discussions on fuel performance, fuel environment (water chemistry), problems with loop operations, and the crud deposition program are included.

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.

Work performed during the quarter is summarized by direct measurement of fission gas pressure, loop operations, performance of UO2 fuel.

The primary purpose of this joint USAEC-Euratom program is to obtain a better understanding of the maximum achievable operating characteristics of UO2 as a reactor fuel. During the program work will be performed in two areas that have been of concern to reactor core designers for a long time, namely, fission gas release and central melting in fuel rods.

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.

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.

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.

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 …

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 …

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.

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.

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

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 …

Experimental investigations were conducted to obtain additional information on the growth and characteristics of sodium oxide deposits in liquid sodium which could lead to system plugging. These investigations included the removal of sodium oxide from molten sodium by the cold trap method.

This study was designed to determine how swaging stainless steel over unground pellets will affect the integrity of fuel rods, and to find maximum pellet diameter variation that can be tolerated.

In order to investigate the low frequency strain cycle fatigue for tubular sheath geometries an apparatus was designed and fabricated for laboratory and reactor experiments. The design of this apparatus is described herein.

The experimental program described has had as its objective the determination of the pressure exerted by free fission gas in operating UO2-filled reactor fuel elements.