The differential cross section for charge-exchange scattering of negative pions by hydrogen has been observed at 230, 260, 290, 317, and 371 Mev. The reaction was observed by detecting one gamma ray from the {pi}{sup 0} decay with a scintillation-counter telescope.

The object of this investigation was to determine an economical method of preparing pure lithium hydroxide from a mercury-contaminated lithium hydroxide monohydrate salt presently produced as a waste product. Pure lithium hydroxide has application for chemical removal of carbon dioxide from air and general open market sale if the mercury contamination is reduced to approximately one part per billion. Because of the uncertainty of the form of the mercury contaminant, different purification methods were explored on a laboratory scale which could be applied to the industrial waste stream. The experimental results indicate that the predominant mercury contaminant existed as mercuric oxide, which was deposited in the by-product salt when the solubility of mercuric oxide, 60 ppm, was exceeded in aqueous lithium hydroxide solution. To purify a fraction of the industrial by-product salt, a crystallization system, utilizing the difference in solubility of lithium hydroxide and mercuric oxide, is proposed. Total stream purification, using sulfide treatment, is expected to be less effective than recrystallization due to the difficulty in physical removal of the mercury contaminant, as mercuric sulfide, from solution.

The 190-KW process water pumping Unit No. 1 is scheduled to be modified for increased pumping capacity under Project CGI-883- Component parts for this modification are expected to-be received during June 1960. Installation of these components would require approximately ten days; due mainly to grinding of the high lift pump case to make room for the new larger diameter impeller. In order to minimize lost production, it has been proposed by K Reactor Operation that the high lift pump be modified early this spring during the scheduled maintenance overhaul period on pumping unit No. 1. The test impeller recently removed from the No. 1 high lift pump in KE would be repaired and installed in the KW pump at this time. Later, in June or July when the components for the complete modification are available the low-lift pump and drive motor would be modified and associated electrical and instrument alterations would be completed during a normal reactor outage. Adoption of the proposed plan would make it necessary to operate the modified high lift pump for a period of approximately three months with an unmodified low-lift pump. A study was made to determine the feasibility of operating the pumping unit in …

Test metal was irradiated to simulate a `blanket` loading and striped E-N load under PT-IP-255-A-9-FP. The product analysis has now been obtained on the test material. The interpretation of this data and plant direction that might follow is considered in this document.

The test was located in the C Reactor. Reactor atmosphere analyses are given. Graphite temperature near the holder was about 500 C at equilibrium operation.

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One of the three major categories of HAPO fuel element failures is the side corrosion type rupture. The majority of side-corrosion failures has been characterized by oval or tear-drop shaped flow patterns containing evidence of accelerated corrosion. Thorough examination of many of these so-called `hot spot` failures has indicated the failure was caused by poor heat transfer associated with misalignment, dimensional distortion or poor jacket-to-core bonding. It has been postulated that misalignment of the fuel element is a necessary condition for formation of hot spots under the present reactor operating conditions. Neither tru-line contours nor X-8001 alloy are successful in the prevention of misalignment and associated ruptures; therefore, it has been proposed to test the effectiveness of projections on the side of the fuel element toward preventing fuel misalignment in ribbed process tubes. A previous test of this element termed the `bumper fuel element` was encouraging; however, it failed to provide the conclusive proof required to justify a large-scale demonstration loading. Supplement A to the basic test was written to obtain necessary preliminary data. This report presents an outline of further testing required to accelerate evaluation of the bumper concept.

The process water flow rates at the 100-K Reactors will be increased as a result of plant modifications under Project CGI-883. The flow rates to be achieved by the project are 188,000 and 200,000 gpm under five and six 190 Building pump operation respectively. Preliminary studies have indicated that the K-Area effluent systems will, for the most part, be adequate for the expanded conditions provided the reactor effluent is continuously and simultaneously discharged from two retention basis to the river by way of the basin overflow weirs. The third basin would be available for any necessary effluent retention or diversion of effluent into crib facilities. The purpose of this document is to present the results of an analytical study of the effluent system capabilities and to set fourth the additional design work with which must be performed in order to complete the detailed design of required alterations to the effluent system.

Production Test IP-272-A-FP authorizes the installation of up to 100 smooth bore Zircaloy-2 process tubes in C Reactor to demonstrate the feasibility of self-supported fuel elements for production use. An additional 200 zirconium tubes are expected to be delivered by mid-year and con be used to expand the initial demonstration facility. It is the purpose of this document to assess the status of the pilot demonstration program from the B-C Reactor Operation viewpoint.

Recent reviews of the last-ditch water backup system at the old reactors had shown that complete adequacy could not be demonstrated at all of the reactors under all conditions. However, these conclusions were based in part on conservative calculations using basic data with a sizeable amount of possible error. Subsequently, recommendations were made to run several tests which would increase the accuracy of the basic data and thereby allow an increase in the demonstratable level of last-ditch vater backup adequacy. The water supply capabilities of the last-ditch system are normally measured with a tripout test where reactor riser pressures are recorded over a period of time while the reactor vater supply is transferred to the last-ditch system. These riser pressure readings are then converted to reactor flows later by means of a reactor hydraulic demand curve, so the inaccuracy in the low-flow region of these curves result in an inaccurate assessment of the water flow obtained from the last-ditch system during the tripout test. The purpose of this document is to report the results of the aforementioned low-flow calibration tests at B, D, F, and H Reactors.

A resume of Health Physics activities for October, November, and December, 1959 is presented. Discussions and tabulations which summarize results of field surveys, bioassay, personnel monitoring, and environmental serveys are included.

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Activities in a program to determine the effects of product size (or dimensions), composition, and heat treatment on beta-treated U are reported. Progress to date was chiefly confined to organizing the program and setting up equipment. A Jominy end-quench apparatus was constructed, and preliminary experiments were run on as-cast metal to test the Jominy-oscillographthermocouple system. Results are tabulated. The method to be used for characterization of beta-treated samples by determination of growth index and texture is also described. (J.R.D.)

The rear face or discharge area of a reactor contains all the appurtenances necessary to discharge irradiated fuel, to collect hot coolant from each process tube, to monitor tube and effluent temperatures, and to monitor the coolant for ruptured fuel elements. Generally, failure of a rear face piping component would not affect the safety of the reactor since the coolant has fulfilled its purpose, that of cooling the fuel elements. The failure may, however, cause failure of one of the monitoring devices and if undetected could lead to a minor reactor incident. The Purpose of this report is to review all information generated during the past three years concerning the condition of rear face piping and hardware. This review includes the history of rear face piping and hardware problems, study activities taken to ascertain the condition of the components, action taken to correct actual component failures, programs recommended to correct deficiencies which operating experience and engineering judgement indicate are necessary, and programs to accumulate additional information to support design of new piping and hardware components.

This document studies a broad study program which is currently in progress in Irradiation Processing Department to evaluate the technical and economic feasibility of various methods of obtaining increased production from the six older reactors (B, D, DR, F, H, and C Reactors). Due to time limitations this study has been in general terms only, but has indicated that considerable increased plant return can be obtained from an increased conversion ratio as well as from higher reactor power levels. The work performed by this Unit has been concerned with defining the reactor process component modifications, and process piping changes between the 105 Building valve pits and effluent basins which will be required to attain the production increases.

gone approximately 16 years of almost continuous irradiation. Throughout this time stored energy has accumulated at a slow rate to the present maximum value of about 35 cal/gm releasable to 250 deg C. A small portion of the moderator (approximately 4%) contains stored energy which under adiabatic conditions may be released spontaneously (at approximately 165 deg C) to produce a maximum temperature of 270 deg C. Careful analysis has shown that the presert condition is not hazardous; however, it appears wise at this time to initiate some corrective action (thermal annealing) to prevent the continued buildup of stored energy to a dangerously high value. Several methode of obtaining effective annealing in the OGR were investigated. The proposed method was selected upon the basis of convenience, over-all safety, effectiveness, and cost. The proposed method involves the alteration of the present coolant flow system to permit reversal of air flow through the fuel channels. This will result in a reversed temperature distribution wherein the maximum graphite temperature will occur in the normally cold, maximum-stored-energy region of the moderator, Such an arrangement permits an annealing operation to be performed under conditions very similar to those of the normal safe operation. The proposed procedure …

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Water depth in the intake channel to the coolant-water screenhouse and surrounding area in the Ohio River near the intake channel were measured. From these measurements, the amount of silt removed during a previous dredging was determined. (C.J.G.)

The potentiometric behavior of zirconium in fluoride, nitrate, and fluoride-nitrate systems is correlated and discussed in terms of the zirconium species present. A potentiometric titration was used successfully for the determination of total nitrate and zirconium in aqueous or organic (tributyl phosphate) zirconium-nitrate systems. (auth)

The admittance and transfer functions of large, solidcore electromagnets were determined. The effects of eddy currents and hysteresis were considered in deriving the functions. The study was concerned with the type of magnet employed in nuclear physics research which requires very precise regulation of the magnetic field. The study originated during the design of an analyzing magnet regulator for the ORNL 63-inch cyclotron. (W.D.M.)

The release of fission products from porous fuel elements during irradiation may be largely a diffusion process. An equivalent-sphere hypothesis was proposed to provide a model by which the diffusion can be analyzed. The equations of diffusion were previously solved for in- pile conditions. The slow convergence of the formulas makes the previous solutions awkward. In the present investigation alternate formulas were derived which are more suitable under certain circumstances. Tables were prepared from which release rates and accumulations may be evaluated for prescribed conditions. The application of the analysis to the interpretation of release data is explained. (auth)

Tensile, bend ductility, and hardness tests were performed at room temperature on irradiated tantalum sheet to determine the effect of irradiation on the strength and ductility. Sheet tensile specimens were irradiated in an attempt to produce corversions of tantalum to tungsten of approximately 1.5 and 3.0 wt.%. Unirradiated tantalum and arc-melted alloys of tantalum-1.5 and -3 wt.% tungsten were tested for comparison with the irradiated material. The tensile and yield strengths of tantalum were found to increase appreciably as a result of irradiation whereas the tensile properties of unirradiated Ta-W alloys prepared by arc melting showed that small additions of tungsten do not signicantly increase the strength of tantalum. These results indicate that the major part of the increase in strength resulting from irradiation of tantalum can be attributed to fast-neutron damage and that any contribution produced by the conversion of tantalum to tungsten is a minor one. (auth)

A developmental program was conducted to provide and in-pile loop facility for use in evaluating gas-cooledreactor fuel asubassemblies. The program included the design, construction, and installation of a recirculating gas loop which is located in a 6 by 6-in. facility in the aluminum reflector of the ETR. The loop system was designed to recirculate the primary nitrogen coolant at flow rates up to 0.9 lb per sec and pressures up to 200 psia. It will accept fuel subassmeblies up to 36 in. in length and 2.26 ia. in diameter with specimen power generation up to 150 kw. The maximum coolant temperature at the specimen outlet is set at 1500 deg F. The loop system includes the in-reactor section, the machinery, the control system, and the specimen-handling apparatus. Salient features of the re-ertrant system include an aluminum pressure wall in the in-reactor section, static gas insulation between the reactor coolant and the circulating loop gas, and a controllable rate of heat exchange between the specimen inlet- and specimen outlet-gas channels in sections of concentric countedlow piping. The three blowers in the system feature grease-lubricated bearings and water cooling. The complete system was tested out of pile and is now installed in …