Oscillations have been observed in a magnetically supported cylindrical rod of plasma. This rod of plasma can be the discharge occurring in the defining aperture of a Mode II, pressure gradient arc. Similar oscillations can also occur in the column of a Mode I arc. These oscillations appear to be the mechanism that drives the Mode II blowup phenomena.

Preliminary results from experiments on the dissolution of beryllium metal and sintered UO2-BeO fuel pellets are reported. In all cases the pellets were fired in hydrogen at 1650-1800°C. Uranium, from UO2-BeO pellets containing more than 60% UO2, is readily leached with boiling 6-13 M HNO3 in about 6 hr. The BeO in these pellets dissolves only slowly in nitric acid; however, in 8 M HNO3-0.2M NaF, it dissolves at about the same rate as the UO2. Sintered pellets containing less than 10% UO2 do not dissolve rapidly in common aqueous reagents. The highest rates are obtained in boiling acidic fluoride solutions; e.g., sintered BeO and BeO-8% UO2 are dissolved initially at a rate of about 1.7 mg min-1cm-2 (13 mils/hr) in boiling 5.8 M NH4HF2.

The absorption in lead of Bremsstrahlung X radiation from a Sr^90 TiO3 pellet in the proximity of Hastelloy "C" was measured. The tenth value layer of the more energetic components of the X-ray continuum was determined to be 1.60 inches.

The General Electric Company is proceeding with an irradiation program to proof test a representative array of Savannah I fuel rods. Irradiation of a test assembly containing Savannah I fuel rods has begun and it is proposed that the results of this irradiation will permit an advance evaluation of the fuel performance and fuel burnup in the Savannah I reactor. This report covers the first two quarters of the reporting period. All aspects of the subject program have been consolidated and applicable portions are discussed in some detail.

This paper reports the chemical and physical properties of the NTA film packet. It correlates with these properties the response of this packet to neutrons of various energies. In this correlation the concept of the track unit is introduced as a basic unit for reporting film-packet response.

Technical report describing and evaluating the the three experimental methods for obtaining equation of state data at low temperatures; (1) approximate measurement of the PVT relationship by a piston-displacement technique, (2) the measurement of a heat capacity at constant volume as a function of molar volume and temperature, and (3) direct measurement of the pressure variation of the elastic constants using ultrasonic techniques. X-ray methods also might be applicable.

None

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.

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.

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.

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)

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

None

None