The effectiveness of purification in controlling plant radioactivation rates was determined through comparison of the actual effects produced from plant operation with and without purification. With the purification system in service, there was no appreciable difficulty in maintaining the reactor coolant within reference water specifications. In addition, there was no discernible increase of crud deposition in the coolant system as determined by direct radiation measurements of the purification hairpin loop. However, without demineralization as a controlling agent, the gross non-volatile gamma activity levels of the reactor coolant increased and the specific activities of longlived fission products were at higher levels. Although the general levels of water- borne activities increased during the test period, the associated plant systems did not exceed their limits. (auth)

EQUIPOISE-3 is an IBM-7090 FORTRAN programmed code for the solution of two-group, two-dimensional, neutron diffusion equations. A maximum of 2l00 mesh points may be used, and the code will solve problems in either rectangular or cylindrical geometry. Logarithmic derivative boundary conditions are allowed, and removal of neutrons from both groups is permitted. Adjoint fluxes with the associated fluxadjoint flux regional integrals may be calculated automatically if desired. A constant buckling, group-dependent buckling, or region-dependent buckling may be specified for rectangular geometry. This program is intended to fill the need for a rapid two-dimensional calculation suitable for survey calculations. During the iterative part of the computations, all operations are carried out in the core memory. The magnetic tape memory is used only for input, output, and program storage. The running time for a 1000-point problem requiring 100 iterations would be about 3 min.(auth)

Activities in research programs are summarized in the areas of power reactor fuel processing, fluoride volatility processing, molten salt reactor fuel processing, homogeneous reactor fuel processing, waste treatment and disposal pilot plant decontamination, GCR coolant purification studies, equipment decontamination, HRP thoria blanket development, fuel cycle development, transuranium element studies, production of U/sup 232/, uranium processing, fission product recovery, thorium recovery from granite, solvent extraction technology, mechanisms of separation processes, radiation effects on catalysts, ion exchange technology, chemical engineering research, chemical applications of nuclear explosions, reactor evaluation studies, and assistance programs. (J.R.D.)

Ring-joint flanges were studied in thermal-cycle tests as part of the development work associated with Homogeneous Reactor Experiment No. 2 (HRE-2). The purpose of this study was to provide criteria for design, installation, and operation of joints that would remain leaktight under reactor operating temperatures and pressures. Joints ranging from 1/2 in., l500 lb to 4 in., 2500 lb and with various initial bolt loadings were cycled between room temperature and 636 deg F. It was demonstrated that when joints were made up to HRE-2 standards and specifications, leak rates of less than 0.25 x 10/sup -3/ g of water per day per inch of gasket pitch diameter could be routinely- attained. Undamaged gaskets could be reinstalled or new gaskets used with equal probability of achieving acceptable leak rates. The system installed in HRE-2 was provided with a high-pressure buffer system to ensure that the small amount of leakage to the cell would be nonradioactive. (auth)

Recovery of Th (and U) from Granitic Rock. Recovery of Th by acid leaching ten addltlonal granite samples (36 to 82 ppm Th) from the Conway formation in N. H. ranged from about 50 to 85%, and averaged about 70%, Study of the effect of grind size on the recovery of Th from Conway and Plkes Peak granites showed no significant differences in the range minus 20 to minus 200 mesh. The Th concentration in a sized Conway granite sample was found to be much greater in the fine than in the coarse fractions, whereas Pikes Peak granite showed only slight Th enrichment in the finer fractions. U recoveries in acid leaching of four different granite samples were not improved by adding an oxidant. Collection and Analysis of Granite Samples. A field survey of the Conway granite formations in N. H. was made. Preliminary analysis of the data indicates that the accessible surface of the Conway granite averages at least 40 ppm Th. Collection and Analysis of Lateritic Soils. The Th concentration ranged 5 to 16 ppm in twenty-two samples of sub-lateritic soil from Miss., Ala., Ga., and Va. Final Cycle Pu Recovery by Amine Extraction. In continued batch countercurrent …

A total of 25 sections is incorporated in the report. Separate abstracts were prepared for 22 of the sections. Those sections for which no abstracts were prepared are concerned with Th fuel cycle development, assistance programs, and publications. (J.R.D.)

The effects of the maximum credible accident relative to the design of the containment shell are discussed. The maximum credible accident is defined. The thermal and hydraulic effects of the maximum credible accident on the reactor system were analyzed. The extent to which fuelrod cladding will melt was estimated. The amount of energy released from the reactor system by the escaping steam and water and by a possible chemical reaction was calculated along with the corresponding pressure rise inside the containment shell. The kinds, amounts, and total radioactivity of fission products released to the atmosphere of the containment shell after the core melts were predicted. (M.C.G.)

We present the results of measurements of the angular distribution of fission fragments produced by irradiation of Au{sup 197} and Bi{sup 209} with various heavy ions. The projectiles, B{sup 11}, C{sup 12}, N{sup 14}, and O{sup 16}, had energies from a few MeV above the Coulomb barrier to 10.4 MeV per nucleon. The gross features of these results can be explained by use of a model and parameters that have been used by others to account for angular distributions of fission fragments from helium-ion bombardments. In detail, however, these results appear to indicate that the models used to predict the average value of the angular momentum of the compound nucleus give values too low near the Coulomb barrier. Furthermore, at high bombarding energies it is necessary to consider the fact that appreciable direct interaction is taking place.

This report addresses the configuration of Tatum Salt Dome, Lamar County, Mississippi.

This report, from the Chemical Processing Department at HAPO for September 1963, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; weapons manufacturing operation; and power and crafts operation.

This report, from the Chemical Processing Department at HAPO for January 1966, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

This report for March 1961, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.

This report, from the Chemical Processing Department at HAPO for October 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

The PRTR is entering a phase of operation in which there is an increased danger of high level fission product release from failed fuel. The higher fuel operating temperatures result in a much larger percentage fission gas release from the fuel crystal; also it is believed that the central voids and inter-crystal cracks greatly improve the transport of fission gases from the far end of a fuel rod to a defect location. 6 tabs.

This report, from the Chemical Processing Department at HAPO for May 1963, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; weapons manufacturing operation; and power and crafts operation.

This report, from the Chemical Processing Department at HAPO for November 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

This report itemizes the irradiations performed by Testing for Battelle-Northwest Laboratories. It lists the material being irradiated, awaiting disposition and material shipped during the report period.

In September of 1965, a program was launched to investigate the cause and to determine appropriate corrective or preventive action for hydriding of Zircaloy process tubes in K Reactors. One phase of the investigation was specifically directed to the measurement of galvanic corrosion currents flowing between the Zircaloy process tube and the aluminum spacers. Production Test IP-807 authorized in-reactor galvanic test measurements in order to secure data relating to the hydriding problem. It is the purpose of this report to document all the galvanic corrosion current data recorded for tests in K Reactors during the investigation of the hydride problem. The galvanic potentials recorded for test measurements in the K Reactor process tubes, identifies the approximate location and magnitude of the corrosion currents generated between aluminum spacers and Zircaloy process tube. The flow of galvanic currents and subsequent release of hydrogen gas can be related to the hydriding of the Zircaloy tubes. The data enclosed in this report indicate a substantial amount of galvanic action in the downstream dummy section of the process tube. The use of certain metal coatings on the aluminum spacers could reverse the galvanic cell polarity and cause corrosion of the process tube which would not …

A CIIIE (enriched) element, one of two elements which failed April 5, 1962 in tube 2955C, was shipped to the Radiometallurgy Laboratory for detailed examination. The element was 21st from the downsteam end. A more severe rupture which occurred in the 19th piece from the downsteam end was believed to have caused the initial rupture indication. After a hot startup a heat cycle developed in the part of the reactor containing tube 2955 increasing the bulk water temperature about 20 per cent. At about the same time the temperature rise was noted, the first rupture indication was seen. The reactor was shutdown 40--45 minutes later, tube 2955 was pushed, and the two ruptures were found. The element was classed as a ``hot spot`` failure. Detailed examination was requested to determine the cause of failure and characterize the nature of any observed corrosion.

This report, for June 1965 from the chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

This report, from the Chemical Processing Department at HAPO for February 1966, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

This document discusses the test program of evaluating the temperature effect of crud film build-up on fuel element heat generating surfaces in a carbon steel system. This program has three phases: Measurement of the effect of film build-up during normal high pH equilibrium operation; measurement of the temperature effect of film build-up subsequent to a loop decontamination; and measurement of the effect of film build-up in the event of loss of pH control.

The purpose of the test described in this report was to determine the effect of production reactor flux on ceramic materials being considered for lining the vertical safety rod channels at the production reactors. This document is being written to describe the test, the results to date, and the work which remains to be done. The materials being evaluated in the test are: Aluminum silicate (Maryland Lava Co.); silicon nitride bonded silicon carbide (Norton Co. ``crystalon``); and aluminum oxide 85% (Norton Co. ``Alundum RA-98``).

In response to specific requests of the AEC and as part of a Combined Operations over-all review, a number of engineering studies have been made of alternative methods for increasing availabiltiy of neptunium at Hanford. The report updates an earlier study in which recycling of both natural and enriched uranium was considered for Hanford. The earlier study showed that recycled natural uranium would provide appreciable gains in neptunium availability but at an excessive cost. Recycle of the slightly enriched uranium streams proved a more economical means of realizing smaller but still significant gains in neptunium production. Subsequent to the earlier report, a feasible and immediately applicable scheme for UFI blending has been conceived demonstrating further advantages for recycling the enriched uranium. Approximately 34 kilograins of supplementary neptunium could be produced at Hanford during the next seven to eight years by upgrading irradiated E-metal and NPR uranium through a blending operation at Oak Ridge (rather than in the diffusion cascades) and then recycling the material through the Hanford reactors. (cf Table 1) Such a scheme would conserve uranium-236 for use as a source of neptunium-237 in the reactors without incurring major capital costs. Oak Ridge Operations personnel have estimated that capital …