One-region buckling measurements that were made on a series of D/sub 2/O- moderated lattices of heavy uranium metal tubes in the Process Development Pile at Savannah River Laboratory are presented. The purposes of these measurements are to provide normalization points for lattice bucklings and to extend the study of natural uranium- D/sub 2/O systems. The dependence of buckiing on the moderator-to-fuel ratio is studied for two types of lattices.

Abstract: New experimental data and literature data are utilized to determine the upper temperature of usefulness of zirconium alloys. Three basic engineering assumptions are used; (1) service life requirements are on the order of four years; (2) tubular fuel cladding for rod-type fuel is considered with a maximum wall thickness of 1.27 cm; and (3) heat fluxes are above 157 watts/cm. The inter-relation of three basic factors, corrosion rate, corrosion embrittlement by hydrogen and oxygen, and strength are considered. An upper limit for an acceptable corrosion rate for long-term service of 1 mg/dm/day is set primarily by the effect of heat-transfer on corrosion. For the best alloys anticipated, this requirement (even without considering transient conditions) limits cladding surface temperatures to less than 540 degree C. Oxygen embrittlement of the alloy substrate by oxide film dissolution is not expected to be a limiting factor. Corrosion hydrogen embrittlement was studied in detail and found to limit acceptable service to cladding surface temperatures of less than 525 degree C for established experimental alloys. Hydrogen embrittlement may not be a limiting factor if alloys corrosion resistant enough to be acceptable above 600 degree C could be developed. Zirconium alloys designed for high strength to …

Report issued by Oak Ridge National Laboratory that describes research and progress at the Chemical Technology Division.

The Heavy Water Components Test Reactor (HWCTR) is used to test prototype fuel elements for power reactors that are moderated with heavy water and fueled with natural or slightly enriched uranium. During the initial critical experiments in the HWCTR, it was observed that there were unexpected variations in nuclear reactivity. Investigations revealed that this effect was due to bubble of helium gas appearing and disappearing in the moderator. An examination of the expected operating conditions of the HWCTR and the solubility of helium in D2O showed that it was possible during normal operation for the helium content of the moderator to exceed saturation and thus for helium to appear as bubbles in the moderator. The possibility of helium bubbles appearing in the moderator because of solubility characteristics was eliminated by modifications to the process system so as to maintain the gas content of the moderator appreciably below saturation.

The Heavy Water Components Test Reactor (HWCTR) is a pressurized reactor, cooled and moderated with D2O, and has the capability of testing fuel assemblies under operating conditions of coolant flow, temperate, and pressure that are typical of those proposed for modern power reactors. The report contains (1) description of the four systems used for failed element detection, (2) discussion of the laboratory analyses of water samples used a as backup for the fuel failure instruments, (3) description of 3 monitors, Cyclic Air Sampling Monitor, Stack Gas Activity Monitor, Health Physics Building Monitors, (4) normal full power activity readings, (5) discussion of the experience during fuel failure.

Summary: Two UO2-filled stainless steel clad fuel rods in which fission gas pressure was measured during irradiation have been subjected to post irradiation examination. Results of free gas analysis and metallographic examination are in general agreement with observed pressures reported previously. Calculated fuel surface temperatures based on extent of fuel recrystallization indicate that in a one-half inch diameter fuel rod with 0.014 inch diametral clearance operated at a maximum heat flux of 531,000 Btu/hr-ft, gap conductance increased with increasing heat flux. An analysis of void configuration indicates that pressure is more sensitive to as-fabricated void volume and changes in this volume resulting from fuel expansion than to fuel central temperature. The decreases in effective void volume with increasing fuel temperatures becomes more significant as initial void volume decreases, and excessive fission gas pressures may be developed in fuel rods operated at high fuel temperatures unless adequate expansion volume is provided in fabrication.

The Heavy Water Components Test Reactor (HWCTR) was built for the Atomic Energy Commission by the Du Pont Company to satisfy a need for fuel testing in the AEC's Heavy Water Power Reactor Program. The reactor was designed to provide a realistic test environment for full size fuel candidates. The report contains sections on (1) the containment building, (2) vertical cross section of the reactor vessel, (3) core layout, (4) low power physics tests and comparison with calculations, (5) rod worths, (6) temperature coefficients, (7) flux shapes, and (8) the operating philosophy of a test reactor.

Summary: The use of dimensionless-parameter frequency response diagrams to determine accuracies of lumped-parameter approximations is demonstrated by two examples: calculation of the heat flux at the surface of a semi-infinite solid due to temperature fluctuations of an adjacent fluid; and the response of a counterflow heat exchanger to inlet fluid temperature perturbations. Dimensionless system parameters make it possible to use general-purpose plots to find the error in particular approximations as a function of the frequency of perturbation. Such plots are directly applicable to control-system stability problems, where the highest frequency of interest is usually apparent.

From introduction: "Evaluations to determine cause of fuel rods breakage following irradiation."

Technical report. From Abstract : "The Sr-SrH2 phase diagram was studied by thermal analysis, chemical analysis of equilibrated phases and X-ray diffraction. The maximum solubility of SrH2 in strontium metal is 38 mole % at the peritectic temperature of 880°C. Strontium metal undergoes an allotropic transformation at 555°C and melts at 768°C. A second transformation was found, at about 240°C, in samples containing hydrogen. Strontium hydride was found to have an allotropic transformation at 855°C."