An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Photoinduced electron spin resonance signals have been observed in isolated chloroplasts and other green materials with a growth time not affected by reducing the temperature to -140 deg. This is interpreted in terms of conduction-band and trapped-electron theory.

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Corrosion studies indicate that digest liquors presently in use at the Mallinckrodt Uranium Refining Center should not be excessively corrosive to the digesters, which are constructed of Type 347 stainless steel. Experimental tests have shown that the digest liquors resulting from some of the more recent sources of uranium concentrates might become corrosive if the fluoride, chloride, or free nitric acid contents of these liquors should reach abnormal levels. If such constituents are present, there are corrective procedures that may be employed to prevent excessive corrosion of the stainless steel. It has been demonstrated that aluminum additions to the digest liquors will reduce the attack in both the liquid and vapor phases when relatively high concentrations of fluoride are involved. Iron additions may be used to some extent to combat corrosion arising from high chloride levels. Type 304 ELC stainless steel appears to have comparable corrosion resistance to Type 347 and Carpenter No. 20 stainless. Haynes Alloy No. 25, which showed a somewhat lower rate of penetration tended to form a loose insoluble corrosion product which may be objectionable in service. (auth)

The elements of waste treatment and disposal are assessed which are expected to become important in the development of the nuclear power industry of the future. Growth of the nuclear power economy is considered along with composition and quantities of anticipated waste. In addition, the economic implications of waste disposal are considered. It is concluded that research should be concentrated on decontaminating off-gases and on conversion of wastes to a more suitable form than liquid for storage. (J.R.D.)

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A rough calculation shows that a flux of 10/sup 16/ n/cm/sup 2/ sec can be obtained in the internal thermal column (island) of a reflector-moderated circulating fluoride-fuel reactor. Existing NaZrF/sub 5/ base fuels and a graphite moderator are used. The average power density in the reactor is 1 kw/ cc, and the total pcwer is 444 Mw. Inner radius of the fuel region is 50 cm. (auth)

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The feasibility of the electronic-explosive safety device was demonstrated in KEWB-I reactor excursions. It was shown to be a fast, efficient safety system for reactor protection. The safety devices inserted 1.5% negative reactivity into the core region of the KEWB-1 reactor to produce rapid shutdown. The system response time, i.e., that time interval between trip and insertion times, was found to be 4.2 msec. The complete freedom in choice of trip level and the flexibility of device geometry makes the system applicable to many types of reactor and critical assembly. The system does not offer the ultimate in reactor safety because it is not completely self-contained, but it is a practical, workable system for reactor protection. (auth)

jectives, the accomplishments, and a summary of the work outstanding. The obtectives of the experimental and analytical studies were to investigate and reach an understanding of the kinetic behavior of aqueous homogeneous reactors. Information produced by the program, experiments on the spherical core, capsule experiments, and the remaining work schedule are discussed. (W.D.M.)

The equations, numerical values, and results of the calculation are given. The assumptions involved and the known uncertainties are listed. (M.H.R.)

A number of techniques have been investigated on a laboratory-scale for separating uranium from fluorides during the recovery of uranium from UF4 reduction bomb wastes (C-oxide) by an HCl leach - NH4OH precipitation process. Among these are included adsorption of fluorides from filtered leach liquors, fractional precipitation of fluorides and uranium, complexing of fluorides into forms soluble in slightly acid solutions, and fluoride volatilization from the uranium concentrate. Solubility studies of CaF2 and MgF2 in aqueous hydrochloric acid at various acidities and temperatures were also conducted. A description of the production-scale processing of C-oxide in the FMPC scrap plant has been included.

This report concerns the continuing problem of providing waste tank designs which will insure safe storage of radioactive waste over long periods of time. These tanks must withstand severe stresses imposed by earth loading, by hydrostatic pressure and by temperature gradients. The temperature gradients are caused by contained in the waste. The waste entering the tank is relatively cold, but the temperature rises rather rapidly to the boiling point as filling proceeds and then the temperature continues to rise as self-concentration increases the boiling point of the waste. The temperature gradients are therefore transient in character and they may be calculated only by rather tedious and complex methods; however, the gradients must be established in order to provide design data for structural design purposes. This report describes the methods used to calculate the temperature gradients and presents the calculated results for a particular tank design used under various operating conditions.

The recovery of fission products from Hanford wastes has for some time been under investigation by various HAPO workers. Flowsheets for the recovery of cesium have been demonstrated, and one for the recovery of cerium is ready for full-level testing. Several tentative flowheets have also been proposed for the recovery of other fission products and of waste plutonium and uranium. The {open_quotes}Integral Flowsheet{close_quotes} developed by Chemical Research Operation is based primarily on the work of G.B. Barton. The present work is a continuation of that begun by Barton. The primary objective has been the recovery of long lived fission products, other than cesium, with particular emphasis on cerium-144 and strontium-90. Secondary objectives of importance include: (1) the isolation of uranium and plutonium into solutions suitable for recovery by recycle into appropriate Purex plant streams, and (2) gathering data that may be useful at some later date on the recovery of the remaining fission products (other than cerium and strontium) should these become valuable. Precipitation procedures were principally considered since idle plant facilities already exist which were designed for this type of process. It is also desirable that the processes developed be compatible with the already demonstrated cesium recovery flowsheet.

A system for concentrating uranyl nitrate solutions was designed and installed in the Thorex Pilot Plant. A total of 16,060 g of uranium was concentrated in the system in 68 batch runs. A total of 14,400 g total uranium (14.180 g U/sup 233/) was recovered as product suitable for shipment. Uranium loss to the evaporator condensate was 0.03% of ihe total uranium processed. The material balance across the system was 98.4%. The average concentration of uranium in the evaporator feed solution was 29 g/liter; the average concentration in the evaporated solution was 298 g U/liter and in the product solution was 199 g/liter. Radiation readings of bottles containing product solutions were taken with a hard-shell cutie pie immediately after each run, and these readings ranged from 35 to 1100 mr/hr. The radiation levels of the bottles of product solution shipped averaged 78 mr/hr. Bottles of product solution reading in excess of 300 mr/hr, maximum allowable for shipment. were reprocessed in the second-cycle solvent extraction system (Thorex) and reconcentrated. The products from seven runs had radiation levels in excess of 300 mr/hr at the time of concentration, or the activities had grown to that level by the time of shipment. The …

A network of 50 stations located at weather bureau installations collected daily samples of settled and airborne dust in the area beyond 200 mi from the test site of operations BUSTER and JANGLE. This network was augmented by mobile teams operaiing immediately beneath the trajectories of the radioactive clouds and by aerial collections made along the 85th and 95th meridians. Data collected by the Weather Bureau stations sented extrapolated to Feb. 1, 1952. The collection of background data prior to the initial test or the BUSTER series was hampered due to two atomic explosions in a foreign country just prior to the start of this program. Particle size determinations by two independent methods revealed that substantially all the activity was contained within the range of respirable particle sizes. (C.H.)