Report describing a zero-power reactor facility (ZPR-VII), its systems, and associated hazards.

Report regarding "[t]he Fuel Cycle Facility for the Second Experimental Breeder Reactor (EBR-II), the process equipment, and the operations to be conducted in the facility are described. The Fuel Cycle Facility is a plant for reprocessing, by pyro-metallurgical methods, the core and blanket material discharged from EBR-II. The reactor core alloy is uranium-5 percent fissium and contains about 46 wt.% Uranium-235 . The blanket material consists of uranium in which plutonium is bred. Core and blanket subassemblies contained in transfer coffins are transferred between EBR-II and the Fuel Cycle Facility, which is in an adjacent building." (p. 13)

From Introduction: "The Fast Reactor Test (FARET) program is part of the United States Fast Reactor Program directed toward the demonstration of breeder reactors essential to the long-range goal of utilizing fertile fuels. (1) With-in this program is the current thinking that the early construction and operation of two or three prototype nuclear power plants of the order of 250 Mw(e) will lead eventually to the construction of practical and economic full scale breeder reactors by the early 1980's. (2)"

Report concerning the study of the two-phase flow through the cooling channels of a natural-circulation boiling water nuclear reactor. "One-dimensional conservation equations describing the flow through each channel are written in the linearized perturbed form, and Laplace transformation in time is performed." (p. 5)

From Abstract: "An alternative approach is presented with attempts to demonstrate that insight may be gained from a consideration that the basic cause of the effect may be due to reactivity rather than to period as the previous authors assumed. Methods for compensating for this effect are discussed."

Report containing "analyses of thin, flat, circular plates subject to bending" (p. 7) using various equations for use as equipment for the Argonne National Laboratory Zero Gradient Synchrotron.

Report containing "analyses of thin, flat, circular plates subject to bending" (p. 7) using various equations for use as equipment for the Argonne National Laboratory Zero Gradient Synchrotron.

"WR16, a program written in 3600-FORTRAN for a CDC-3600 digital computer, calculates, as a function of time, the concentrations of all the species taking part in the simultaneous first- and second-order chemical reactions occurring during and after exposure to ionizing radiations occurring during and after exposure to ionizing radiation of any chemical system (for example, an aqueous solution). Homogeneous reaction kinetics is assumed for the computation. The variation of electrical conductivity and optical absorbance of the system can also be computed" (p. 1).

Report concerning "the unstable behavior of a heat-transfer loop operating at a super-critical pressure (p. 7)" within a nuclear reactor.

Report issued by the Argonne National Laboratory discussing acoustic-boiling-detection techniques. As stated in the abstract, "this report summarizes literature through June 1967 concerning acoustic methods. In the acoustic method for boiling detection, either acoustic waveguides or high-temperature acoustic sensors are recommended" (p. 1). This report includes tables, and illustrations.

Report issued by the Argonne National Laboratory discussing meteorological data collected between 1952 and 1953. Wind, temperature, pressure, and precipitation studies are presented. This report includes tables, illustrations, and a map.

Report issued by the Argonne National Laboratory discussing the PACER-65 computer program. As stated in the summary, the program "has been developed and utilized for flow sheet calculations and process-data reduction. PACER-65 is an executive program in which material- and energy-balance equations, conversion factors, etc., for each processing step are described by separate subroutines" (p. 5). This report includes tables, and illustrations.

Report issued by the Argonne National Laboratory discussing progress made by the Reactor Development Program during March 1963. Reactor physics, experiments, and safety studies are presented. This report includes tables, and illustrations.

Report issued by the Argonne National Laboratory covering the quarterly report from the Reactor Engineering Division. A summary of reactor programs, designs, development, and experiments are presented. This report includes tables, illustrations, and photographs.

Report discussing the relative thermal conductivities of liquid lithium, sodium, and eutectic NaK, and the specific heat of liquid lithium, as well as the methods and materials used to determine this information.

A set of OS/370 Basic Assembly Language programs is described which provides a FORTRAN IV interface with OS/370 Macros.

Vapor explosions were produced by injecting small quantities of water into a container filled with molten sodium chloride. Minimum explosion efficiencies, as evaluated from reaction-impulse measurements, were relatively large. Subsurface movies showed that the explosions resulted from a two-step sequence: an initial bulk-mixing phase in which the two liquids intermix on a large scale, but remain locally separated by an insulating gas-vapor layer; and a second step, immediately following breakdown of the gas layer, during which the two liquids locally fragment, intermix, and pressurize very rapidly. The experimental results were compared with various mechanistic models that had been proposed to explain vapor explosions. Early models seemed inconsistent with the results. More recent theories suggest that vapor explosions may be caused by a nucleation limit or by dynamic mixing combined with high surface-heat-transfer rates. Both types of models are consistent with the results.

Progress report describing the research and management efforts of Argonne National Laboratory's program on high-performance lithium/metal sulfide batteries during the period July-December 1975. The batteries are being developed for two applications: off-peak energy storage in electric utility networks and electric-vehicle propulsion. The battery designs for the two applications differ, particularly in cell configuration and electrode design because of the differing performance requirements.

A simple conduction model with phase change has been developed for the transient analysis of a fuel pin based on average properties and lumped-parameter techniques. The transient behavior of fuel and cladding can be accurately described by simple analytical expressions that agree with conventional numerical approaches for under-cooling transient analysis. If it be assumed that the heat-transfer resistance between the fuel and cladding remains the same for both steady-state and transient periods, the phase-change problem for fuel and cladding melting can be significantly simplified. BY using the predetermined average overall heat-transfer coefficient across a fuel pin in the steady-state period, the average transient fuel and cladding temperatures can be formulated analytically. For loss of flow at constant power, the start of melting and complete melting for both the fuel and cladding can be estimated with considerable accuracy.

The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed.

Hydrogen from the radiolysis of water by dissolved fission products is stripped from the solution and collected by bubbling carbon dioxide through the solution. Quantitative measurements of the G value for hydrogen show that the yield is essentially the same as would be obtained by external gamma radiolysis of nonradioactive solutions of the same chemical composition. The hydrogen yield can be enhanced by addition of a hydrogen-atom donor, such as formic acid, to the solution. The yield of hydrogen from fission-waste solutions is discussed with respect to the question of whether it represents a significant energy source.

Studies of unprotected loss-of-flow accidents in the CRBR for various rates of flow coast-down and with various options in the SAS 3A code did not lead to conditions for a violent disassembly. Maximum fuel temperatures using the SLUMPY module for disassembly were in the range 4000-4500 deg C. An approximate treatment of the LOF-driven TOP accident, not properly modeled by SAS 3A, indicates the possibility of some increase in accident severity. The effect of fission gas in dispersing fuel was not taken into account in these calculations. Parameter variations included the presence or absence of axial fuel expansion and of clad motion and use of the moving coolant film model versus the static film model. Study of severe pipe rupture accidents with scram indicated that pin power density and fuel-clad conductance were important parameters in determining what coolant flow rate was needed to prevent boiling after the rupture. It appears that for the CRBR when engineering hot channel factors are considered, this fraction would have to exceed 25 percent.

Current status of epidemiological studies of the late effects of internal radium in humans, and mechanistic investigations of those effects.

Quarterly report on activities of Argonne National Laboratory's Nuclear Technical Programs, including results of studies to measure the degradation of backfill materials after their exposure to temperature and humidity expected in high-level nuclear waste repositories.