The computer programs, including HYBRID, VSYS, VEHIC and CRASH, used to compute the energy and fuel consumption, life-cycle costs and performance characteristics of a hybrid electric-powered vehicle are described and their use documented. (LCL)

The design requirements, design philosophy, method and assumptions, and preliminary computer-aided design of the Near-Term Hybrid Vehicle including its electric and heat power units, control equipment, transmission system, body, and overall vehicle characteristics are presented. (LCL)

The UDD sensor bar with modified cross-section (0.0938-in. thick by 0.25-in. high overall) was analyzed to determine its resonant frequencies. Thermal and fluid-induced stresses due to steady-state and LOCE loads were also calculated. The sensor bar was shown not to be subject to vortex-shedding lock-in and was shown to meet applicable criteria of the ASME Boiler and Pressure Vessel Code. Seismic loads are insignificant compared to fluid loads.

The results of Task I, mission analysis and performance specifications, for the Near-Term Hybrid Vehicle (NTHV) development program are presented. The items researched included trip characteristics, market potential, life-cycle cost and performance specifications of NTHV's. (LCL)

This Phase I summary report contains a description of the 8 kW wind energy conversion system developed by the United Technologies Research Center (UTRC) for the Department of Energy. The wind turbine employs the UTRC Bearingless Rotor Concept in conjunction with a passive pendulum control system which controls blade pitch for start-up, efficient power generation, and high-speed survivability. The report contains a summary of the experimental and analytical programs in support of design efforts. These supporting programs include materials tests, a wind tunnel program, and aeroelastic analyses to evaluate system stability. An estimate is also made of the projected manufacturing cost of the system if produced in quantity.

Near stoichiometric bulk polycrystalline CdSiAs/sub 2/ has been synthesized by two techniques: (1) direct fusion of the elements and (2) direct fusion of the binaries SiAs, Cd/sub 3/As/sub 2/ and CdAs/sub 2/. The latter technique resulted in denser ternary material with good homogeneity. The above binaries melt congruently and were also formed by direct fusion. Sputtered ternary films were formed using a bulk CdSiAs/sub 2/ target, and a composite target of CdAs/sub 2/ discs in a Si plate. Composition of the CdSiAS/sub 2/ target changed with sputtering time. Amorphous films deposited from that target were heat treated, and became crystalline and near stoichiometric but with poor mechanical properties. It appears that films deposited from the composite target (Si + CdAs/sub 2/) can be adjusted to stoichiometry by means of sputtering power and target geometry. As deposited, these films also were amorphous. With respect to evaporated films, the study of thermal decomposition of CdSiAs/sub 2/ in vacuum was completed. The decomposition is preferential toward Cd between 570/sup 0/ and 710/sup 0/C, and toward As in the 710 to 1010/sup 0/C range. It is concluded that evaporation of the ternary is not a suitable method for forming CdSiAs/sub 2/ films. Plans for …

The Composite Bearingless Wind Turbine was developed at the United Technologies Research Center following many years of bearingless rotor research and development for helicopter applications. The current 8 kW contract was begun in October 1977 with delivery of the prototype scheduled for November 1979. The contract consists of two phases. Phase I encompasses the complete design of the system and supporting analytical and experimental studies. The Phase I final report has been completed, and this executive summary highlights the principal findings described in that report.

Research is reported on materials for magnetic fusion energy, laser fusion energy, Al-air batteries, geothermal energy, oil shale, nuclear waste management, thermochemical cycles for hydrogen production, chemistry, and basic energy science. (FS)

The CORA code is a steady state/transient, core thermal hydraulics code for the FFTF Reactor. A brief overview of the code development and use is presented.

The purpose of this joint ORNL/Westinghouse Program is to develop a design concept for a tokamak reactor blanket system which satisfies engineering requirements for a utility environment. While previous blanket studies have focused primarily on performance issues (thermal, neutronic, and structural), this study has emphasized consideration of reliability, fabricability, and lifetime.

Recent studies have shown geothermal power plants to have a significant environmental impact on the ground water of the area. The heavy metals arsenic and mercury are special problems, as both are concentrated by flora and fauna exposed to the effluent waters. Because the toxicity of these and other metallic pollutants present in geothermal effluent depends on the chemical form, or speciation, of the particular metal, any serious study of the environmental impact of a geothermal development should include studies of trace metal speciation, in addition to trace metal concentration. This proposal details a method for determining metal speciation in dilute waters. The method is based on ion-exchange and backed by atomic absorption spectrometry and multiple scanning anodic stripping voltammetry. Special laboratory studies will be performed on mercury, arsenic and selenium speciation in synthetic geothermal water. The method will be applied to three known geothermal areas in Washington and Oregon, with emphasis on the speciation of mercury, arsenic and selenium in these waters. The computer controlled electrochemical instrumentation was built and tested. Using this instrumentation, a new experimental procedure was developed to determine the chemical form (speciation) of metal ions in very dilute solutions (ng/ml). This method was tested on …

In March 1979, ETEC published as ETEC-TDR-78-19 a search which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This Supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions was found, nor were any newer data on saturated or subcooled conditions uncovered. The Supplement also updated a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of ETEC-TDR-78.19.

Documentation of the extreme aridity (approx. 3% relative humidity) of the lower stratosphere and the rapid decrease of mixing ratio with height just above the polar tropopause (20-fold in the 1st km) was begun by Dobson et al., (1946) in 1943. They recognized that this extreme and persistent aridity must be dynamically maintained else it would have been wiped out by turbulent diffusion. This led Brewer (1949) to hypothesize a stratospheric circulation in which all air enters through the tropical tropopause where it is freeze dried to a mass mixing ratio of 2 to 3 ppM. This dry air then spreads poleward and descends through the polar tropopauses overpowering upward transport of water vapor by diffusion which would otherwise be permitted by the much warmer temperatures of the polar tropopauses. Questions can indeed be raised as to the absolute magnitudes of stratospheric mixing ratios, the effective temperature of the tropical tropopause cold trap, the reality of winter pole freeze-dry sinks and the representativeness of the available observations suggesting an H/sub 2/O mixing ratio maximum just above the tropical tropopause and a constant mixing ratio from the tropopause to 30 to 35 km. However, no model that better fits all of …

A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

The MARK I boiling water reactor (BWR) containment system is comprised of a light-bulb-shaped reactor compartment connected through vent pipes to a torus-shaped and partially water-filled pressure suppression chamber, or the wetwell. During either a normally occurring safety relief valve (SRV) discharge or a hypothetical loss-of-coolant accident (LOCA), air or steam is forced into the wetwell water pool for condensation and results in hydrodynamically induced loads on the torus shell. An analytical program is described which employs the finite element method to investigate the influence of torus wall flexibility on hydrodynamically induced pressure and the resultant force on the torus shell surface. The shell flexibility is characterized by the diameter-to-thickness ratio which is varied from the perfectly rigid case to the nominal plant condition. The general conclusion reached is that torus wall flexibility decreases both the maximum pressure seen by the shell wall and the total vertical load resulted from the hydrodynamically induced pressure.

One organization's approach to designing complex, interrelated databases is described. The problems encountered and the techniques developed are discussed. A set of software tools to aid the designer and to produce an initial database design directly is presented. 5 figures.

The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear fuel rod elongation and as an indication of critical heat flux and the occurrence of departure from nucleate boiling. These types of measurements provide important information about the behavior of nuclear fuel rods under normal and abnormal operating conditions. The objective of the paper is to provide a complete account of recent advances made in LVDT design and experimental data from in-core nuclear reactor tests which use the LVDT.

In designing a plutonium-uranium core with thorium internal blankets there are several parameters that must be investigated to identify the design with the best performance. Choices include: the amount of moderation in the blanket, whether moderation is incorporated from the outset or added after an assembly is shuffled to the external blanket, and the cycle length both for the core and internal blanket assemblies. To investigate the effect of these parameters several core and blanket burnup analyses were performed.

Recent developments in computer based systems for quantitative x-ray microanalysis, 4 Pi surface examination, Auger electron spectroscopy and Backscattered Microtopography measurement have extended the Scanning Electron Microscope's applications in ICF target development and production.

This progress report for the USNRC/LLL Liaison program with the Federal Republic of Germany regarding boiling water reactor containment multivent steam condensation tests being conducted by GKSS (Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt) address program activity during the period of May-June 1979. During this period the program scope was defined, initial contacts between LLL and GKSS were established, and the first trip by an LLL representative to the GKSS test facility was taken.

Kinetic models for the reaction between a gas and powder are presented and discussed. Rates from the progressive conversion model are dependent on the amount of material present. This model is described by an equation that yields both the rate constant and the reaction order. The phase boundary model produces rates for topochemical reactions which are dependent on the powder surface area. Applications of these models to uranium powder hydriding are discussed. Computer programs for analyzing data also are included. 6 figures.

The multiplication of 14 MeV D-T fusion neutrons via (n,2n), (n,3n), and fission reactions by /sup 238/U is well known and established. This study consistently evaluates the effectiveness of a depleted (tails) UO/sub 2/ multiplier on increasing the production of /sup 233/U and tritium in a thorium/lithium fusion--fission hybrid blanket. Nuclear performance is evaluated as a function of exposure and zone thickness.

A design is presented that suggests that a negative ion neutral beam based on direct extraction is applicable to TNS, assuming technological advancements in several areas. Improvements in negative ion sources, direct energy conversion of charged beams, and high speed cryogenic pumping are needed. The increase in efficiency over a positive ion system and the encouraging results of the first attempt at a total design justify increased effort in the development of the above mentioned areas.

An ignition tokamak reactor design has been obtained which represents a starting point for the conceptual design of a 1000 MW/sub e/ commercial system. The design includes Nb/sub 3/Sn superconducting coils (TF, OH, and SF), water-cooled fissile blanket (e.g., uranium oxide), positive-ion based neutral beams with no direct energy recovery, and an ignited plasma with a = 0.9 m and an aspect ratio A = 4.0. The TF coil bore has a vertical bore of 7 m and a horizontal bore of 5 m both of which are a factor of two larger than the corresponding bore dimensions of the LCP (Large Coil Project) TF coil. The plasma is characterized as follows: stability factor q = 2.5, Z/sub eff/ approx. 1, poloidal beta ..beta../sub p/ less than or equal to A, a elongation delta in the range between 1.6 and 1.7. A number of potential operating conditions for the plasma and device have been identified for which the plasma beta ..beta.. lies within the range from 6.5% to 7.3%, and the plasma temperature has an average value between 11 keV and 12.5 keV. The design was obtained using the computer code COAST and represents a self-consistent sizing and costing result.