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.

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.

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."

Abstract: The question of how crack-free, protective oxide films can form on zirconium during oxidation when the Pilling-Bedworth ratio is about 1.5 has been considered by a study of the relative plasticity of various forms of zirconia. Hot hardness measurements showed that doping mono-clinic zirconia with iron, nickel, or chromium resulted in softer (more plastic) structures and that yttrium additions slightly reduced the plasticity. Calcia-stabilized cubic zirconia was found to be more plastic than mono-clinic zirconia when tested at temperatures above 200 degrees C. The behavior of anion-deficient oxides indicated that they were more plastic than stoichiometric oxides even though the hardness values were identical at 23 degrees C. The former were free from cracks at the indentions, whereas, stoichiometric oxides exhibited extensive cracking around and between indentions. The behavior of actual, thick (72 microns) oxide films during tensile deformation of oxidized metal samples indicated that considerable plasticity occurs in the oxide at 500 degrees C but that the films are brittle at 23 degrees C. It was concluded that the plasticity of the oxide may be greater than that of the oxygen-contaminated substrate at elevated temperatures and may be the means by which epitaxial strains are minimized.

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

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.

Abstract: This technical report analyzes gamma dose rate and neutron measurements in their relation to the SM-1 shield design and is a continuation of previous shielding measurements and analyses reported in APAE-35 and APAE-35 Supplement 2. The data reported herein are spent fuel element and rod drive pit gamma dose rates. An analysis of gamma dose rates off the core midplane is presented and compared with test data.

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.

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.

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.

From introductory paragraphs: "The central problems in neutron research are the understanding of nuclear structure and the study of the properties of nuclear particles, particularly the properties of the neutron. The most fruitful attack on these problems is the determination of the probability of interactions between neutrons and nuclei, i.e., the measurement of neutron cross sections. Ideally, this involves the study of all possible types of neutron interaction with all available nuclei at all neutron energies...The discussion in this paper will omit the interactions leading to neutron productions, and will be limited to the intersections of neutrons with stable nuclei."

Procedures for the separation and concentration of microgram to nanogram amounts of thallium from gram amounts to galena and meteorite metal, sulfide, and silicate phases were developed and are described. The thallium is extracted from aqueous bromide of chloride solutions of moderate acidity using diethyl ether. A device for elimination of the large volumes of solvent by dropwise evaporation is described. Organic matter in the residue is destroyed by treatment with aqua regia, and the thallium is the residue is converted to the nitrate for spectrochemical or mass-spectrometric examination. The presence of thallium in the residue is tested by Rhodamine B. test.

"A number of mixtures of polymers and multi-unsaturated monomers have been prepared and irradiated. The content of insoluble gel of irradiated samples of these polymer-monomer mixtures has been determined by extraction."