A discussion is given of possible septum magnet parameters for small production angle experiments in the ISABELLE storage rings. Superconducting septa and torroidal septa are also considered. (PMA)

A simple analytic method is developed, based on physical approximations, for treating transient implosive longitudinal compression of bunches of heavy ions in an accelerator system for ignition of inertial-confinement fusion pellet targets. Parametric dependences of attainable compressions and of beam path lengths and times during compression are indicated for ramped pulsed-gap lines, rf systems in storage and accumulator rings, and composite systems, including sections of free drift. It appears that for high-confidence pellets in a plant producing 1000 MW of electric power the needed pulse lengths cannot be obtained with rings alone unless an unreasonably large number of them are used, independent of choice of rf harmonic number. In contrast, pulsed-gap lines alone can meet this need. The effects of an initial inward compressive drift and of longitudinal emittance are included.

This document contains copies of viewgraphs discussed in a presentation made at the Fifth Annual Power Conference, August 29 to September 2, 1977. No text. 19 figures, 11 references.

Time-dependent deformation and time-dependent strength are being characterized for several candidate polymeric composites for flywheels. This presentation highlights the motivation and the philosophy of the characterization adopted by the authors in establishing the ongoing programs at LLL. This overview is intended to provide a basis for inferring the type of enginering data being generated for different aspects of flywheel design. The details of these data can be obtained from the published reports and articles. Two aspects of flywheel design data are addressed: those dealing with time-dependent statistical strength, and those dealing with deformation and strength under time-varying history.

The past, present, and future thrusts of the magnetic mirror fusion program at LLL are reviewed. Neutral beam injection, stabilization, and density-lifetime product results from the 2XIIB experiment are briefly highlighted. The rationale of the Tandem Mirror Experiment and Field Reversed Mirror Experiment now under way are discussed. Plans for the Mirror Fusion Test Facility (MFTF) are described. Approaches to improvement of particle containment in mirror fusion systems are briefly indicated. (RME)

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.

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 …

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.

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.

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.

The feasibility of conducting fusion blanket integral neutronics experiments at the Rotating Target Neutron Source-II (RTNS-II) accelerator facility and the Tokamak Fusion Test Reactor (TFTR) was investigated. RTNS-II recently became operational, and the TFTR is scheduled to begin D-T operations during 1983. The experiments would provide reaction rate data of direct importance to blanket design in environments (neutron spectra) close to those expected in actual blankets. Data of this kind are especially important for a hybrid blanket, where design depends upon a balance of breeding and power production requirements. The blanket also provides an essential part of the toroidal field (TF) coil shielding. Therefore, experimental verification of design model calculations is important before any commitment to a definitie design is made.

Initiation and growth of a crack in the Charpy V-notch test was investigated by performing both static and impact controlled deflection tests. Test specimens were deformed to various deflections, heat-tinted to mark crack extension and broken apart at low temperature to allow extension measurements. Measurement of the crack extension provided an estimate of crack initiation as defined by different criteria. Crack initiation starts well before maximum load, and is dependent on the definition of ''initiation''. Using a definition of first micro-initiation away from the ductile blunting, computer model predictions agreed favorably with the experimental results.

A /sup 1///sub 5/-scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system.

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm/sup 2/) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm/sup 2/) and for ISX-B/sup 2/ (11 kA/cm/sup 2/). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure.

Plasma confinement and neutral beam injection heating were investigated on the Oak Ridge Tokamak (ORMAK) plasma with improved plasma parameters due to higher injection power (to 360 kW), discharge current (to 220 kA) and toroidal field (to 26 kG). With increasing injection power up to 360 kW with otherwise constant operational parameters, the central ion temperature increased roughly linearly from 0.7 keV to 1.8 keV. The scaling of ion temperature with injection power and plasma density reasonably agrees with theoretical predictions based on neoclassical ion heat conduction and classical beam energy transport.

Tritium containment and removal problems in a fuel-reprocessing plant are identified and conceptual process designs for reducing emissions to the environment to below 1 Ci/day are studied. The conceptual design recommended would allow an air atmosphere in the reprocessing-plant hall and would use a continuous-catalytic-oxidizer/molecular-sieve-adsorber cleanup system to maintain a 40-..mu..Ci/m/sup 3/ tritium level (5 ..mu..Ci/m/sup 3/ HTO) against 180 Ci/day leakage from components and process piping.

Two ''plugs'' of dense plasma at either end of a central solenoid cell form the basis of a new mirror fusion power plant concept. A central cell blanket design is presented. Modules on crawler tracks serviced by remote welding and handling machines of very simple design are important features resulting from linear axisymmetric geometry. Three blanket designs are considered and the best one presented in some detail. It has lithium as the breeder material, helium cooled. ''Plug'' magnet field strengths must be high. A novel magnet is presented to satisfy the physics of the end plugs. Beam sources at 1,200 KV present special problems. Methods of voltage standoff, arc damage control, and neutralization are discussed. New secondary containment ideas are presented to allow removable roof sections of balanced design.

Vacuum pumping by freezing out or otherwise immobilizing the pumped gas is an old concept. In several plasma physics experiments for controlled fusion research, cryopumping has been used to provide clean, ultrahigh vacua. Present day fusion research devices, which rely almost universally upon neutral beams for heating, are high gas throughput systems, the pumping of which is best accomplished by cryopumping in the high mass-flow, moderate-to-high vacuum regime. Cryopumping systems have been developed for neutral beam injection systems on several fusion experiments (HVTS, TFTR) and are being developed for the overall pumping of a large, high-throughput mirror containment experiment (MFTF). In operation, these large cryopumps will require periodic defrosting, some schemes for which are discussed, along with other operational considerations. The development of cryopumps for fusion reactors is begun with the TFTR and MFTF systems. Likely paths for necessary further development for power-producing reactors are also discussed.

Tritium-containment systems for the blanket and power systems of a mirror-fusion reactor are described. These systems are designed to reduce emissions to below 1 Ci/d. The overall conceptual design uses air as the reactor-hall atmosphere. A continuous catalytic oxidizer-molecular sieve adsorber cleanup system would be used to control a 180-Ci/d leakage from reactor components, energy recovery systems, and process piping. Such a system would maintain a 40 ..mu..Ci/m/sup 3/ tritium level (5 ..mu..Ci/m/sup 3/ HTO) in the hall. The blanket considered contains submodules with Li/sub 2/Be/sub 2/O/sub 3/-Be for tritium breeding. This canned breeding material is scavenged with a lithium-vapor-doped helium gas stream. The container consists of molybdenum alloy (TZM) tubes and tube sheets with the breeding material packed and sintered in the shell surrounding the tubes. Potassium vapor coolant (also lithium-doped) passes through these tubes to recover the heat at 950/sup 0/C. Leakage following an intermediate TZM exchanger would result in a loss of 0.7 Ci/d into the steam through the Haynes-25 alloy boiler (potassium boiling). A moving getter bed is used to recover the tritium from the LiT and Li/sub 2/T scavengers in both the helium blanket scavenging flow and the potassium vapor coolant.