Originally issued as S and P 1963A, Parts I and II. This report covers a design and feasibility study of a pebble bed reactor-steam power plant of 125 megawatt electrical output. The reactor design which evolved from this study is a two-region thermal breeder, operating on the uranium-thorium cycle, in which all core structural materials are graphite. Fuel is in the form of unclad spherical elements of graphite, containing fissile and fertile material. The primary loop consists of the reactor plus three steam generators and blowers in parallel. Plant design and system analysis including cost analysis and capital cost summary are given.

Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.

Abstract: Accurate measurement of OMRE fuel-element surface temperatures is important for evaluation of the performance of organic compound as moderator and coolant materials in nuclear reactors; a mixture of polyphenyls has been used in reactor operation to-date.

Work is reported in these areas: general radiation effects; toxicity of radioelements; studies at the molecular and cellular levels; environmental radiation studies; atmospheric radioactivity and fallout; marine sciences; radiological and health physics; and radiation instruments.

Progress is updated for projects related mostly to biological and ecological radiation effects.

Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.

"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."

Abstract: The structural design criteria and the system basic loads for the SNAP 2 compact power unit are presented.

Abstract: Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of commercially available valves considered for use in the sodium coolant system of the Hallam Nuclear Power Facility.