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.

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.

Methods used to control the radiological impact of the nuclear fuel cycle are described. This control is exercised through the application of a series of federal laws and regulations that are used as the basis for licensing nuclear facilities. The control is exercised more directly by the use of radwaste treatment equipment at the nuclear facilities to limit the release of radioactive materials. Federal laws and regulations are summarized and their applications in licensing actions are discussed. Radiological doses from materials released from licensed facilities are compared with doses from natural background. A series of cost/benefit engineering surveys are being made to determine the cost and effectiveness of radwaste systems for decreasing the release of radioactive materials from model fuel cycle facilities and to determine the benefits in terms of reduction in dose commitment to individuals and populations in surrounding areas.

The principles of mirror machine confinement are reviewed with emphasis on the physical process of neutral beam injection and plasma end leakage. The characteristics of efficient neutral beam injectors and direct energy convertors for the plasma and leakage are described.

A review is made of the current status of /sup 235/U fission cross section data from thermal to 20 MeV neutron energies. The accuracy achieved is compared with the 1 percent accuracy required of a reaction-cross-section standard throughout this range. The energy ranges from thermal to 10 keV, 10 keV to 0.8 MeV and 0.8 to 20 MeV are considered separately because of the different experimental techniques required in each. The goal of normalizing all fission cross sections to the thermal value and the current degree of success is discussed.