This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR.

A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX.

The use of large, functional mockups as a means to develop fusion reactor remote maintenance technology was examined. Critical remote maintenance requirements were categorized and one example, the removal and replacement of a torus sector, was used to illustrate the use of mockups. A generic mockup system was developed and the engineering issues which could be addressed were listed. Consideration was given to techniques involved in designing and using functional mockups for this purpose.

This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated.

Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ignition-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a four-rail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were unplugged and returned to a large, centrally located hot shop. A similar concept is envisioned for compact fusion reactor testing.

Development of blanket designs for fusion reactors requires module testing at reactor-relevant power densities, which cannot be obtained on near-term devices without augmented heating. Concepts are investigated which will substantially increase fusion blanket bulk heating over that produced by incident flux from a fusion or fission reactor, thus permitting meaningful near-term blanket testing on lower-power fusion devices. Electrical heating is considered and limitations determined for its use. A heat enhancement technique for bulk heating blanket materials is also proposed which uses the high energy released by fission of /sup 235/U. Calculations indicate that doping the blanket module with low percentages (2 to 3 vol %) of /sup 235/UO/sub 2/ will provide heat enhancement exceeding a factor of five. Calculated results of heating profiles for Li/sub 2/O and LiAlO/sub 2/ blankets are presented.

Design requirements, trade studies, design descriptions, conceptual designs, and cost estimates have been completed for the Fusion Engineering Device sector handling machine, movable manipulator system, subcomponent handling machine, and limiter blade handling machine. This information will be used by the Fusion Engineering Design Center to begin to determine the cost and magnitude of the effort required to perform remote maintenance on the Fusion Engineering Device. The designs presented are by no means optimum, and the costs estimates are rough-order-of-magnitude.

The overall objective of this task is to investigate the extent to which the power density in the FED breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow testing of advanced reactor blanket modules on INTOR at representative conditions. The tentative approach adopted for this task consists of three parts. First, the requirements for augmented heating of the test module are outlined for different applications of interest. Second, methods are identified which have potential for augmenting the heating power in a test module, and this list of methods is narrowed to those which appear to be most useful. Finally, these methods are examined in more detail to determine the practical benefits of employing each.

Maintenance equipment which is needed to remotely handle fusion device components is being conceptually developed for the Fusion Engineering Design Center. This will test the assumption that these equipment needs can be satisfied by present technology. In addition, the development of equipment conceptual designs will allow for cost estimates which have a much higher degree of certainty. Accurate equipment costs will be useful for assessments which trade off gains in availability as a function of increased investments in maintenance equipment.

The Fusion Engineering Device (FED) is being designed in a torus shape using ten removable segments to form the torus geometry. The torus consists of a frame and ten shield assemblies which fit into the frame and are held in place structurally using electrically insulated backing plates. It is then necessary to seal the shield segment to the frame for the assembly to sustain an internal vacuum of 10/sup -7/ torr. This task is intended to be accomplished by welding a frame seal between the frame and the shield segment. An example of this concept is shown. This document covers the equipment requirements and pre-conceptual design description for installing and removing the frame seal.

The Fusion Engineering Device (FED) is being designed in a torus shape using ten removable segments to form the torus geometry. The torus consists of a frame and ten shield assemblies which fit into the frame. It is necessary to seal the shield segment to the frame for the assembly to sustain an internal vacuum. Designs for the seal, the welder to weld the seal in place, the cutter to remove the seal, and the handling fixture for seal installation and removal are presented. The concept for the seal installation is novel in that precise alignment of the seal to the torus frame and shield assemblies is not required. Vacuum handling technology is used for handling the three story tall, twelve foot wide, fragile, relatively light weight seal. The welder and cutter assemblies track off the seal handling fixture, eliminating the need for complex rails on each of the shield segments. The entire seal installation and removal system has been designed for remote operation.