Separate abstracts were prepared for thirty-one of the thirty-three sections. Of the sections not abstracted, the one entitled Fission Product Transport'' contained no information, the other, Transport of Noble Gases in Graphite'' is available in a more complete form as ORNLTM-I35 (NSA 16: 9209) (J.R.D.)

The id Thorex process, in which nitric acid is the salting'' agent in the solvent extraction of Th and U from an acid eficient feed with TBP in Amsco, was demonstrated in laboratory equipment for the recovery of synthetic Co solidated Edison Thorium Reactor fuel containing trace fission products. The acid was removed from solution of the declad fuel pellets to produce an acid deficient feed by steam stripping, and the adjusted feed was treated with bisulfite to decrease the extractability of fission products. The Th and U were extracted with 30% TBP in Amsco, and pregnant organic was scrubbed with dilute nitric acid to improve decontamination. Additional acid was added to the extraction section to increase the distribution oefficient of the Th, The Th and U could then be stripped ther simultaneously or separately. No difficulties were ound with either excessive reflux of acid or <0.3%. The co centrated aqueous waste was about 0.2 liter per kilog am of Th processed, about 1/10 of that from the aluminum n trate-salted process. Engineering studies showed that the stage height in the extraction column with the acid Thorex process was 2.1 ft compared with 4 ft with the aluminum-salted Thorex process. …

A review of radiation effects on graphite is presented. Included are discussions of the general relations of these effects with original structural properties, and details of radioinduced changes. Other discussions are devoted to stored energy, annealing, and future problems in the use of nuclear graphite. Data and illustrations concerning structure and radioinduced changes are included. (J.R.D.)

A number of organisms have been examined for their ability to produce electron-spin-resonance signals at low temperatures in response to illumination. The efficiency of the response is of the order of not less than 5%, and the wavelength for maximum response is generally slightly on the longer side of the wavelength of maximum absorption, with a minimum appearing at the wavelength of maximum absorption.

The rail shipment of large quantities of radiocesium involves a potential accidental release of this material in a readily available form to the biosphere. The magnitude of the associated potential damage to man and his environment is evaluated in this report. The evaluation of the consequences of an accidental release of Cs-137 from the Shielded Transfer Tank, Model II (STT) assumes loss of Cs-137 to the atmosphere or to surface-water. Release to the atmosphere could result from a collision followed by fire or explosion. In the event of a fire, a small fraction of the Cs-137 vould be volatilized. An explosion would disperse the Cs-137 still adsorbed to Decalso as particulates. In either case, the material is assumed to be dispersed by atmospheric mechanisms which can be described by modified Sutton equations. The accident involving a fire or explosion assumes that 1 percent or 10 percent, respectively, of 90,000 curies of Cs-137 is dispersed in a metropolitan area. Contamination of the surrounding suburban area is also involved. Damage estimates amount to about 60 million dollars and 400 million dollars, corresponding to a 1 percent and a 10 percent release respectively. Another possible type of accident involves the release of the …

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The objective of this supplement is to determine the effects of operation with neutral pH water on crud deposition on fuel surfaces. This supplement authorizes testing designed to show the effects of coolant pH on the thickness of the crud layer deposited on fuel elements. The supplement will not require additional reactor down time nor will it introduce any hazards to the reactor.

This report presents the technical observations and conclusions to date of the full-reactor demonstration E-N loading for the combined production of plutonium and tritium. A major incentive for such a loading is the increase in conversion ratio resulting from three factors: (1) A ``blacker`` lattice,, in which a combination of ``E`` metal fuel elements (uranium enriched to 0.95% U{sup 235}) and ``N`` metal target elements (enriched lithium-aluminum alloy) approximately matches the reactivity properties of a natural uranium lattice, results in a smaller proportion of parasitic capture of neutrons in the moderator and other reactor components. (2) The capture of neutrons in fringe target material results in the formation of useful product by a significant fraction of the neutrons otherwise lost by leakage. (3) The burnout of product atoms (plutonium and tritium) is reduced in the blacker E-N lattice, thereby increasing the net yield of product atoms per MWD of heat generation compared to the natural loading. Potential advantages for plant-wide E-N production include the routine processing of a single metal stream for Hanford, increased product quality at reduced throughput, and increased production and recovery of higher isotopes in recycled streams.

The primary effort to date in connection with this study has been directed toward obtaining source data which indicates (1) the functions performed during reactor outages and the distribution of time required to accomplish these corrective functions, (2) the groups of crafts associated with each of the recovery functions performed, and (3) the radiation exposures experienced during these activities. The first phase of preliminary analysis has been based on the ``time accountability`` report data originated by the various reactor analysts. The attached computer tabulation is one of the analyses performed considering the time and date a reactor was shut down, the ``cause`` for which it went down and the time and date the reactor was considered back on-line. The report summarizes these accountability data into the following summaries in the order presented below: (1) Total hours down per reactor per cause. (April, 1961 to April, 1962) (2) Number of records indicating experience of outages per reactor per cause. (3) The average and standard deviation; same relationship. (4) Outage summary; total hours down, percentage contribution to the department total outage, and time operating efficiency. (5) Department summary (self explanatory). (6) through (21). Interval between like outages by cause. These reports illustrate …

At the request of B-C Maintenance Engineering, this study was initiated to investigate the feasibility of replacing the 190-C and 105-C process water flow-meter orifices with venturi tubes. Specific aspects of the problem studied were potential flow increases, accompanying production increases, costs, and the ability of the existing 190-C pumps and motors to provide the potential flow increase.

This report provides the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials.

This report describes the procedures and techniques involved in area reduction measurements of fractured tensile specimens.

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An evaluation of two thin-walled outer tubes showed that more extensive alpha working of the billet core stock results in more uniform cladding on the extruded tube. In an effort to eliminate breakthrough and to reduce eccentricity, shift, and bending of the mandrel, two experimental coppernickel billets with Zircaloy sleeves were extruded to check a modified billet design. lt was observed that the final grain size of the unalloyed uranium core of a thin- walled outer tube is insensitive to small variations in the cooling rate from the beta-treatment temperature. An axia' load of 3000 pounds applied to a thin- walled outer tube during autoclaving was ineffective in preventing bowing of the tube. Shipping experiments demonstrated that current packaging methods of thin- walled inner tubes do not prevent bowing during transit. The fabrication of specimens for the capsule irradiation program was concluded with the shipment of sixteen specimens and excess extruded tube stock to Savannah River Laboratory. The following core compositions were represented: U-1 wt.% Si, unalloyed dingot uranium, U-0.3 wt.% Al-0.5 wt.% Si, and U-0.3 wt.% Cr-0.3 wt.% Mo. All irradiation specimens were supplied in the beta-treated condition. The mechanical behavior of Zircaloy-4-clad dingot uranium tube sections was evaluated …

The effects of core barrier coatings, void spaces, and surface-cleaning techniques on the quality of Zircaloyclad flat-plate UO/sub 2/ fuel elements prepared by gas-pressure bonding were investigated. Techniques were developed for the application of barrier layers of chromium by a vapordeposition process and of crystalline carbon by a pyrolytic process. These coatings, together with a graphite coating previously developed, were evaluated in pressure-bonded fuel elements for their effectiveness in preventing coreto-cladding reaction and for their general production feasibility. Crystalline carbon coatings 15 to 40 mu in. thick and chromium coatings 25 to 40 mu in. thick were determined to be satisfactory. In the stady of the flow of cladding-plate material into void spaces in the picture-frame assembly, it was established that excessive flow, and consequent thinning of the cladding, can be minimized by individually compartmentalizing the cores with Zircaloy ribs. This design resulted in maximum restriction of the effects of a cladding failure in service. Quantitative data on the maximum amount of void space resulting from manufucturing tolerances or from chipped fuel cores that is tolerable in cladding failure in service. Quantitative data on the maximum amount of void space resulting from manufucturing tolerances or from chipped fuel cores that …

This report addresses the cryogenic mock-up loop design study.

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This report presents details of the status of special reactor process tube loadings for the month of May 1966.