The fallout hazard and protection factors in current use for large groups of buildings, i.e., urban residential areas, business districts, industrial complexes, government centers, Atomic Energy Commission facilities, and academic and medical institutions, are largely unsubstantiated by experimental evidence. These data are important for personnel protection on a national basis in the event of war and on a local basis in the event of certain types of nuclear accidents. The need for such information is discussed and methods for obtaining it are suggested. The methods suggested should provide a cross check of the data obtained on isolated structures under actual fallout conditions with the data from studies that made use of methods such as distributed point sources and a moving single-point source (as used in the Mobile Radiological Measurement Unit, Civil Effects Test Operations) to simulate actual fallout fields and with data from other studies in which predicted values of fallout protection were calculated from strictiy theoretical considerations. (auth)

The reactor design which forms the base for the current economic status of D/sub 2/O-moderated reactors was estimated from developments in several reactor programs. However, since a heavy water-moderated reactor was not operated on natural U fuel at power reactor conditions, considerable improvement from this current status can be foreseen. A summary of improvements is presented concerning the concept which would result solely from operation of succeeding generation plants without a parallel development program, and improvements which would result from the successful completion of the development program as presented. One plant size was used in the evaluation of plant potential, with a 300 Mw/sub e/ nominal rating. The boiling D/sub 2/O-cooled, pressure tube direct cycle plant design was used. The current development program is outlined; this work includes several items leading to the long-range development of the concept. (auth)

An evaluation of 300- and 100-Mwe power plants was conducted using ground rules prescribed by the USAEC for this study. Costs corresponding to two average discharged fuel burnups are: 8.6 mills/kwh (8500 Mwd/ metric ton) and 8.8 mills/kwh (7500 Mwd/metric ton) for the 300-Mw plant. Costs for the 100 Mw plant are 14.7 mills/kwh for an average discharged fuel burnup of 6010 Mwd/metric ton. Estimates of future potential indicate that the 300 Mw/sub 3/ (8500 Mwd/metric ton) plant could produce power for 7.3 mills/kwh in a second generation, full scale plant of the same type. A further reduction to 6.4 mills/kwh should be possible as the result of the recommended ten-year development program. The current development program is adequate for providing the data needed to design and construct a prototype reactor. However, there is no natural U-fueled prototype and no prototype of the chosen reference design scheduled in the U.S. Current technology is sufficiently developed to initiate the design and construction of a pressure tube, boiling D/sub 2/Ocooled, natural UO/sub 2/- fueled reactor prototype plant in the immediate future. This plant would demonstrate the main features of a full scale plant and, in addition. would provide design data which could …

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Charts are given which present information on the properties, sources, and uses of Th/sup 230/, Th/sup 228/, and U/sup 232/ They also compare these properties with competitive isotopes. (M.C.G.)

The Salt Cycle Process is a nuclear fuel processing approach designed for application in compact facilities located at the reactor sites. Irradiated UO/sub 2/ fuels would be processed through a brief sequence of steps and partially purified UO/sub 2/ or UO/sub 2/--PuO/sub 2/ powders recovered, suitable for refabrication into fuel elements. The major steps of the process are the dissolution of uranium oxides in molten NaCl--KCl eutectic by chlorination to form soluble uranyl chloride and the reduction of uranyl chloride to UO/sub 2/, which separates as a solid. The preferred method of reduction is the electrolytic method, which yields UO/sub 2/ as a conveniently handled, adherent deposit on the cathode. Means are described for separation of Pu from U and for co-deposition of the Uo. Also included are discussions of the probable nature of the U and Pu species present in the molten salt, side reactions in which uranyl chloride may participate, and the influence of certain variables on the properties of the UO/sub 2/ produced. (auth)

The design information is presented in sections on core thermal performance, fuel elements, moderator, and control rods. (J.R.D.)

BS> A method for the measurement of thermal conductivities of electrically conducting materials at temperatures in excess of 1000 deg C with a high accuracy was devised. The spatial boron concentration gradient in boron- doped graphite to achieve maximum p-type thermoelectric output was established. The production of an n-type silicon carbide-graphite composit was studied. Theoretical studies produced a detailed picture of the effect of doping on the electronic properties of graphite. A method was developed for the preparation of rare earth nitrides in a good state of purity. Electrical studies of the materials produced showed that several of the compounds are semiconductors as had been predicted. (M.C.G.)

Quarterly report describing progress on a project to develop and evaluate high-temperature tungsten alloys. This report discusses experimentation with fabrication to improve thermal stability of tested combinations by further alloying.

Abstract: A definitive investigation of the boron-carbon equilibrium system is being made by X-ray diffraction, metallographic, and thermal analytical techniques. Alloys are being produced by sintering pressed powder aggregates with subsequent arc melting. Alloys have been made at two atomic percent intervals up to thirty atomic per cent carbon. In the future, higher carbon compositions are to be investigated. Techniques have been worked out for the metallographic preparation of the extremely hard and friable alloys.

This progress report by the Atomic Energy Division of the Savannah River Plant covers: Reactor Technology; Separation Technology; Engineering Assistance; Health Physics; and General Laboratory work. (JT)

Materials and specifications for the construction of the N-Reactor buildings are presented.

The objective of this irradiation is to further verify the corrosion rate of tubular-type fuel elements under conditions of high specific power and central core temperatures. This fuel will be the inner tube only of an NPR fuel assembly. As in previous tests, this inner tube rupture will be used to further substantiate the rupture detection instrumentation that is being used in the development of the NPR. Previously unirradiated fuel will be used in this test. The reactor is to operate at full power during the test. Permission is requested for charging two tubular elements The top element will have attached to it a hydraulic mechanism for opening a defect in the outer surface of the tube. The second or bottom element, will be used as a heater element to maintain loop temperature.

Increased interest in the civilian, space, and military applications of isotopic power has prompted a study by which the Chemical Processing Department can provide the interim strontium-90 requirements on a schedule considerably accelerated from that previously proposed. This document summarizes a study of the technical and operational feasibility and hazards involved in: recovery of semi-refined megacurie quantities of Sr{sup 90} from current Purex waste; lag storage of the Sr{sup 90} fraction in the 244 CR Process Vault for Sr{sup 89} decay; and subsequent reconcentration and refinement to a bulk product suitable for isolation and packaging by Hanford Laboratories.

The objective of the production test detailed in this report is to compare the dimensional stability characteristics of fuel elements with alloyed low hydrogen dingot cares and standard fuel elements with ingot cores. The basic measurements of dimensional stability are the average warp and the tube-filling capacity values of the fuel elements.

Abstract: This report summarizes data developed at the Instrument and Controls Division, Oak Ridge National Laboratory on low noise amplifiers used with solid state detectors.

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The objective of this test is to measure the emergency reactor flow through the high pressure crosstie line (HPCT) after the removal of the flow limiting orifices in the HPCT. The flow limiting orifices in the HPCT were removed on July 5, 1960, as per Design Change No. 383. The removal of the flow limiting orifice allowed increased emergency flow and brought the crosstie coolant flow more nearly into conformance with the coolant supply reliability criteria. The purpose of this test is to measure emergency flow under certain conditions so that available flow under all conditions may be more precisely determined.

Abstract: This report presents the vapor pressures versus temperature data graphically for the rare earth elements, yttrium and scandium.

Preliminary studies have, in general, indicated the need for modifications and improvements to the reactor last ditch coolants systems in order to provide adequate safety of operation at power levels programmed for the future. These studies have indicated the need for improved reliability as well as increased capacity for the last ditch coolant systems. A design study is being prepared by Reactor Modification Design to define the scope of the modifications required to provide adequate last ditch systems for the older areas. Adequate last ditch cooling will be provided for the 100-K Areas under Project CGI-844 which is currently in progress. The purpose of this document is to set forth the operating conditions and objectives on which the study will be based.

The purpose of this document is to establish the design scope for the proposed modifications to the reactor ``last ditch`` cooling systems in the 100-B, D, F, H, DR, and C Areas. The objective in making these modifications is to provide adequate ``last-ditch`` reactor coolant flows for safety of operation at power levels currently programmed for the period CY 1964 when additional ``last-ditch`` cooling facilities are planned in connection with major plant modifications. Additional interim modifications may be required for the last ditch system at the 100-C and DR Areas and for the export water system prior to major plant modifications during CY 1964--1965.

The purpose of this report is to present laboratory data concerning thermal and hydraulic changes which occur in a low power process channel during a flow interruption. The experiments were conducted in the Heat Transfer Laboratory of Thermal Hydraulics Operation.

This document details activities of the Fuels Preparation Department during the month of July 1960. (FI)

Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Currently, Al-Si bonded fuel elements show evidence of spire bond separation, and to a lesser degree, can bond separation following irradiation. Such evidence has aroused concern for the ability of the currently produced Al-Si bonded fuel elements to withstand future reactor operating conditions. Several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary margin of safety for good bond layer integrity. Before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. This report presents the design of a test to more fully evaluate Al-Si bond integrity under anticipated future reactor operating conditions.