A design and operational description is presented of a BF/sub 3/ withdrawal mechanism planned for installation on one of the two startup channels at the SM-1. An analysis of possible malfunctions is included. (J.R.D.)

The sorption of the ions of uranium, copper, and nickel on hydrous zirconium oxide was investigated at temperatures from 25 to 250 deg C. The experiments were performed by equilibrating 5 ml of the test solution with 0.5 g of zirconium oxide in a titanium autoclave, which was heated by means of a rocking furnace. The sorption of uranium was affected by characteristics of the zirconium oxide, temperatare of equilibration, and concentrations of uranium and of free acid in the uranyl sulfate solutions. Conclusions are drawn concerning the relationship between each of these factors and uranium sorption. (auth)

A titanium pump loop, designed to circulate aqueous solutions at temperatures and pressures up to 370 deg C and 3000 psia, was constructed. It is to be used to study the chemical stability of uranyl sulfate fuel solutions of interest to the Fluid Fuels Reactor Program. The total loop voluime was minimized so that about 2 liters uf solution was sufficient for loop operation. The equipment includes a sampling system to remove solution samples from the loop while operating at elevated temperature and pressure; a hydroclone to separate and remove any solids and/or heavyphase material formed; and provisions for installation of corrosion test specimens in the main loop stream. All equipment performed satisfactorily at design conditions in tests with water. (auth)

Progress is reported on research and development work in pot calcination and radiant-heat spray calcination studies of synthetic Purex high-level wastes; and sorption studies using synthetic minerals and resins as well as natural minerals. The calcination studies are discussed in terms of batch calcination, melting of pot calcination products, spray calcination, and off-gas treatment; and sorption studies in terms of mineral reactions, fixation chemistry, and condensate wastes. (B.O.G.)

Two methods, the transpiration and direct P/sup 32/ counting methods, were tested for use in measuring the vapor pressure of TBP. With the transpiration technique, measured TBP vapor pressures were sufficiently reproducible, and probably of sufficient accuracy, to warrant improvement of the apparatus and its use in measuring the vapor pressure of TBP over various TBP- diluent-HNO/sub 3/-H//sub 2/O-UO/sub 2/ (NO/sub 3/)sub 2 solutions. With this method the vapor pressure of TBP containing 0.2 wt% H/sub 2/O was determined to be ~0.8 mu while that of TBP saturated with H/sub 2/O, ~6.5 wt% H/sub 2/O, was ~ 0.5l mu . The direct P/sup 32/counting technique was abandoned because of experimental difficulties. (auth)

A compilation is presented, in the twenty-five group Reactor Nuclear Data Tape format, of neutron cross sections for elements of major interest for GE- ANPD reactor analysis. The tabulated data are a reproduction of neutron cross section information contained on the Reactor Nuclear Data Tape, which was recently prepared. A brief outline of methods used in processing of the cross sections is also included. (auth)

None

This study responds to the requests for data on the subject contained in the D.S. Burrows letter entitled ``Processing Plutonium Scrap`` (symbol FAc:EAS) dated June 30, 1961 and received for reading by Chemical Processing Department Management on July 5, 1961. Cognizance is also taken of the July 5 teletype from G.F. Quinn requesting an appendix updating the content of our document HW-63560 (Pu Reclamation -- January 18, 1960) with regard to safety aspects covered therein. Appendix A of this paper answers this request.

This report summarizes the evaluations of the maximum monetary liability associated with cesium-137 shipments in the STT Casks and strontium-90 shipments in the HAPO-IA Cask. These evaluations are of most immediate importance since these shipments are planned for the month of March 1961. These liability evaluations concern the direct consequences of a release of the fission product shipment in each case, but do not cover the probability of such events occurring. The liability evaluation of the STT shipment of cesium-137, 90,000 curies per car, was based upon the following premises. Cesium is a relatively volatile material, and the entire shipment quantity can be vaporized and released as a cloud during involvement of the car in a fire of sufficient magnitude and duration to rupture the cask. This mechanism would create, by far, the greatest potential exposure of personnel and property to the fission product.

None

The purpose of this report is to recommend variable goal plans for natural and enriched bumper fuel elements, specifically for 0-III-NB and 0-III-EB materials, to be irradiated at the B, D, DR, and F Reactors. The average goal exposure for all bumper fuel elements at D Reactor was specified to be 900 MWD/T, with provisions being made for revision by normal procedures. Exposures for enriched bumper material at the B, M. and F Reactors were not specified in the PITA supplement authorizing charging of this material.

It is the objective of this production test to authorize the use of PT IP-216-A (Quality Certification) fuel columns and supplementary numbered columns are required for obtaining routine axial flux measurements suitable for metal performance analysis.

None

This document from the principal engineer of the manufacturing section at Hanford calls attention to irreconcilable costs for two fuel fabrication studies. Specifically, the major trouble areas appear to be in refinery, green salt and machining steps.

None

None

A solvent extraction process was developed to produce high-purity Sr/sup 90/ from an irradiated U reprocessing waste solutlon. The extractant is D2EHPA diluted with TBP and Shell Spray Base. The process uses an acetic acid-acetate huffered aqueous phsse which is countercurrently contracted with the D2EHPA organic phase. Calcium and some Ce/sup 144/ extract with the Sr/sup 90/; extraction of other contaminants (Zr/sup 95/, Nb/sup 95/, Ru/sup 106/, and inert lead and iron) is prevented by adding either DTPA or EDTA to the feed solution. Decontamination from Ca and Ce/sup 144/ is provided by back extraction of the Sr/ sup 90/ into an aqueous 1M citric acid solution. The process was used to isolate and purify about one megacurie of Sr/sup 90/ for subsequent use in the fabrication of thermoelectric power generators as part of the Systems for Nuclear Auxiliary Power (SNAP) program. (auth)

Numerical calculations of energy distributions of ions and electrons in DCX in steady state after burnout are discussed. The d-c plasma potential establishing quasineutrality is also determined. The electron distribution is found to be Maxwellian, but the ion distribution is not. For typical DCX conditions, the plasma potential is an appreciable fraction of the injected ion energy (tens of kev). The code employed, ORACLE code EDDIE, which solves coupled, steady-state Fokker-Planck equations governing the ion and electron distributions in energy space, is discussed. (auth)

None

The test data indicate that the response of the Intermediate-Range log- level-current instrumentation was linear between 5 x 10/sup -//sup 1//sup 1/ to 5 x 10/sup -//sup 7/ amperes for start-up rates of 0.2 to 1.0 decsdes per minute based on the as sumption that the reactor trsnsient behavior can be predicted by the reactor kinetic equation. A comparison of the intermediate range start-up rate readings with the start-up rates determined from the slope of the intermediate-range log-level current as a function of time curves indicates that the intermediate-range start-up rate circuitry was properly aligned. The power inception point, determined from the intermediate-range log-level current and reactor coolant-temperature data, occurred at an average log-level current of 5 x 10/sup -//sup 7/ amperes, assuming that all of the heat of the pumps was disposed of by venting steam to atmosphere. The decade overlap of the intermediate range into the source range varied from 0.95 to 1.95 source range decades with an average overlap of 1.1 decades. The hot-to-cold attenuation, ratio of a source- range flux level (cps) at a primary coolant temperature of 500 deg F to source- range flux level (cps) at a primary coolant temperature of 135 deg F, …

A plugging meter designed for the Hallam Nuclear Power Facility was tested. Plugging performance was satisfactory and excellent reprcducibility was achieved. Although the time limitations precluded complete testing of the automatic control equipment, the tests performed gave satisfactory indication of correct performance. (auth)

Radiation-induced contraction as a function of temperature was studied in needle-like and conventional coke graphites from 450 to 1200 deg C. Fast neutron exposures ranged up to 3.2 x 10/sup 21/ nvt, E) 0.18 Mev, corresponding to approximately 20,000 Mwd/at. in a Hanford graphitemoderated reactor. The results significantly extend information on the effect of high temperature irradiation on needle-like coke graphites manufactured with particle size, impregnation, and graphitization temperature as variables. In general, the transverse contraction rate goes through a broad minimum between the temperatures of 850 and 950 deg C with the least contraction occurring at approximately 800 deg C. Needle coke and CSF graphites show the same effect but the contraction of needle coke graphites is approximately 0.8 that of CSF at all temperatures. In the parallel direction, the rates of contraction are the same for both CSF and needle-coke graphites. (auth)

None

The operation of self-actuated pressurizer steam relief valve of the Shippingport PWR was tested within a prescribed setting of 2300 plus or minus 50 psig. The valve popped at 2265 psig and reseated at 2190 psig, with a time lapse of 12.9 seconds between popping and reseating. Leak rates before and after popping of the valve were 0.70 gallons/ hour and 1.35 gallons/hour, respectively. The valve performed reliably in that it popped within the prescribed setting and reseated with no observable valve chatter or flutter. (auth)