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Appendix 8, Decay of Cerium-144

Description: As part of an earlier program of investigation in this laboratory, studies were made of the gamma ray spectrum and the beta ray spectrum of cerium-144. In the present work, seme coincidence studies were made on one of the beta groups appearing in the cerium-144 decay and on the gamma rays appearing in the deexcitations from the energy levels of praseodymium-144. Sources of cerium-144 were prepared frcm carrier free radioactive cerium-144 as supplied by the Oak Ridge National Laboratory. The sample material was more than two years old at the time of preparation of sources. No additional chemical purification was attempted. Sources for use in the beta crystal spectrometer were mounted on thin Formvar film on spectrometer ring mounts. The gamma ray spectrum of cerium-144 in the energy range 20 kev to 180 kev is shown in Figure 1. This spectrum was determined using a 2-inch by 2-inch NaI(Tl) crystal. The pulse spectrum was analyzed by a Radiation Instrument Development Laboratory (RIDL) 200 channel analyzer. The spectrum gives clear evidence of gamma ray peaks at 34 {+-} 3 kev and 134 {+-} 2 kev. A rather broad peak at 80 kev is observed. An indication of a gamma ray group of energy near 100 kev is also shown. The resolution of the detecting assembly was 9.8 percent at 662 kev. The uncorrected relative intensities of the three groups of 34, 80 and 134 kev are 95, 35, and 100, respectively. These intensities are for the gamma radiation exclusive of internal conversion. Gamma-gamma coincidence measurements were made using two of the 2-inch by 2-inch sodium iodide crystals placed at 90 degrees to one another, Ganmma radiation of a particular energy was selected by means of a single channel analyzer and the 200 channel analyzer was used to analyze any coincident pulses ...
Date: June 1, 1960
Creator: Sathoff, H. J. & Azuma, T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

EML Surface Air Sampling Program, 1990--1993 data

Description: Measurements of the concentrations of specific atmospheric radionuclides in air filter samples collected for the Environmental Measurements Laboratory`s Surface Air Sampling Program (SASP) during 1990--1993, with the exception of April 1993, indicate that anthropogenic radionuclides, in both hemispheres, were at or below the lower limits of detection for the sampling and analytical techniques that were used to collect and measure them. The occasional detection of {sup 137}Cs in some air filter samples may have resulted from resuspension of previously deposited debris. Following the April 6, 1993 accident and release of radionuclides into the atmosphere at a reprocessing plant in the Tomsk-7 military nuclear complex located 16 km north of the Siberian city of Tomsk, Russia, weekly air filter samples from Barrow, Alaska; Thule, Greenland and Moosonee, Canada were selected for special analyses. The naturally occurring radioisotopes that the authors measure, {sup 7}Be and {sup 210}Pb, continue to be detected in most air filter samples. Variations in the annual mean concentrations of {sup 7}Be at many of the sites appear to result primarily from changes in the atmospheric production rate of this cosmogenic radionuclide. Short-term variations in the concentrations of {sup 7}Be and {sup 210}Pb continued to be observed at many sites at which weekly air filter samples were analyzed. The monthly gross gamma-ray activity and the monthly mean surface air concentrations of {sup 7}Be, {sup 95}Zr, {sup 137}Cs, {sup 144}Ce, and {sup 210}Pb measured at sampling sites in SASP during 1990--1993 are presented. The weekly mean surface air concentrations of {sup 7}Be, {sup 95}Zr, {sup 137}Cs, {sup 144}Ce, and {sup 210}Pb for samples collected during 1990--1993 are given for 17 sites.
Date: November 1, 1995
Creator: Larsen, R.J.; Sanderson, C.G. & Kada, J.
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Partner: UNT Libraries Government Documents Department

Further studies on the recovery of fission products and uranium from Purex 1WW

Description: 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.
Date: January 28, 1958
Creator: McKenzie, T. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

LOCALIZATION OF CERIUM-144 IN THE SKELETAL TISSUES OF FETAL RATS

Description: The developing fetal skeleton of the rat was used to study the microscopic localization of the rare earth Ce/sup 144/ in bone. Nineteen-day-old rat fetuses were injected with Ce/sup 144/ in isotonic sodium citrate via the umbilical vein. They were sacrificed one-haif hour after the injection by immersion in 80% alcohol fixative. The fetuses were embedded in paraffin and sectioned at 10 to 12 mu . Semiserial sections were treated with periodic acid- Schiff stain for mucopolysaccharides, and by the von Kossa technique for calcium. Contract x-ray-film autoradiographs and NTB stripping-film autoradiographs were prepared. Photographs of representative sections and their respective autoradiographs are presented for various bones. This material strongly suggests that the initial binding site of Ce/sup 44/ in the skeleton is in the organic matrix of cartilage and bone, especially where this matrix is just entering the calcifiable state. This conclusion is highly dependent upon the specificity of the von Kossa stain for calcium. (auth)
Date: October 18, 1957
Creator: Asling, C.W.; Johnston, M.E.; Durbin, P.W. & Hamilton, J.G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

THE RECOVERY AND PURIFICATION OF MULTI-KILOCURIE QUANTITIES OF FISSION PRODUCT STRONTIUM BY CATION EXCHANGE

Description: In seven production runs, 75,000 curies of Sr/sup 90/ were isolated and purified in the ion-exchange equipment of Hanford Laboratories High Level Cells. The production goal (60,000 curies of purified Sr/sup 90/) and the time schedule were met or exceeded and the product exceeded the customer's purity requirements. The hot-cell-purified strontium made possible the completion on time of the Weather Bureau power source and the Martin 10 watt SNAP VII-A and VIl-C units. The isotopic purity of the strontium product was 56% Sr/sup 90/. The chemical purity was greater than 95% strontium. The Zr-Nb/sup 95/ contamination of the product was less than 5 x 10/sup -5/ curies per curie of Sr/sup 90/, and the Ce- Pr contamination was less than 1 x 10/sup -5/ curies per curie of Sr/sup 90/. The final run yielded 16,500 curies of Sr/sup 90/ in an eight liter product solution. The product was >98% strontium and contained less than one curie of Zr- Nb/sup 90/ or Ce-Pr/sup 144/. Eleven days of continuous operation were required to complete the run. (auth)
Date: September 1, 1961
Creator: Bray, L.A.; Lust, L.F.; Moore, R.L.; Roberts, F.P.; Smith, F.M.; Van Tuyl, H.H. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

RADIOISOTOPE FUELED AUXILIARY POWER UNIT. Quarterly Progress Report No. 7, July-September 1958

Description: Progress made in the development of SNAP-1 and -3 is reported. SNAP-1 development reported includes: boiler development, fuel development, properties of cerium dioxide, materials corrosion, power conversion system development, shielding analysis, hazards evaluation, and ground test development. SNAP-3 development includes: power conversion analysis, thermoelectric generator development, and fuel element development. Information is given on the handling and transportation equipment for SNAP-1. (N.W.R.)
Date: October 31, 1963
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

SNAP I Radioisotope-Fueled Turboelectric Power Conversion System Summary, January 1957 to June 1959

Description: The SNAP I development program was initiated to develop a 500-watt turboelectric power conversion system for space applications, Superheated mercury vapor was used as the heat conversion working fluid. The conversion system was to obtain thermal energy from the decay of a radioisotope fuel such as Ce/sup 144/ . Each of the major components and systems is summarized with respect to initial design objectives, development progress to the point of program termination, results obtained from tests and, where indicated, future growth potential. Reference is made to 10 other reports which describe, in detail, the major components of this power generating system. Also included is a bibliography of documented reports that are related to the power conversion system design criteria or system integration into a flight vehicle. (auth)
Date: June 1, 1960
Creator: Dick, P. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Aerodynamic Re-Entry Analysis. Task 2. Thermoelectric Generator Summary Report

Description: An analytical trajectory and aerothermodynamic analysis of a satellite containing a Task 2 thermoelectric generator was completed. A 300-statute mile circular polar orbit was used for this analysis and the launch was assumed to be from Vandenberg Air Force Base. Results of this study show that upon natural decay from a successful mission, the radio-cerium fuel will burn up in space at high altitude, thus only a very minor amount of radio cerium will be released to the stratosphere. A complete analyses of the fate of the radio-cerium fuel following various aborted launching attempts also was carried out. Charts summarizing the various assumed failures and locations of the fuel following failure are shown. A technical discussion of the methods used in performing the analysis is included in the report. (auth)
Date: December 27, 1960
Creator: Oehrli, R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Chemical Processing Department monthly report, June 1959

Description: Production of Pu from separations plants and output of unfabricated Pu exceeded commitments. Purex plant set a new record high for U processed. Production and shipments of UO{sub 3} met schedules. Purex solvent extraction battery performed below normal, probably because of poor solvent quality. NaOH addition to Redox coating removal waste is being reduced. A 3fold improvement in Recuplex product Al impurity was achieved by means of a specific gravity difference > 0.15 between dilute aqueous feed and extractant. Sintered, high-silica crucibles are being tested in RMA production line in Finished Products Operation. Scope design of a fission product shipping cask was completed; powder temperature should be below 440 F for 1 MCi cerium-144 + impurities. Feasibility of using one outside Purex canyon entrance (stairwell opening) for relief damper opening was tested and found to be insufficient. A drawing of the 6-inch continuous centrifuge being evaluated as a vacuum drum filter on RMA button line was reviewed. Casks were designed for the NPR project. (DLC)
Date: July 22, 1959
Creator: MacCready, W. K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

IMGA [Irradiated Microsphere Gamma Analyzer] examination of the Set No. 4 fuel under project work statement FD-20

Description: Results of an examination of over 10,800 unbonded fuel particles from three irradiated spherical fuel elements by the Irradiated Microsphere Gamma Analyzer system are reported. The investigation was initiated to assess fission product behavior in LEU UO{sub 2} TRISO-coated fuel particles at elevated temperatures. Of the three spheres considered, one was reserved as a control and the other two were subjected to simulated accident-condition temperatures of 1600{degree}C and 1800{degree}C, respectively. For the control sphere and the sphere tested at 1600{degree}C, no statistical evidence of fission product release (cesium) from individual particles was observed. At fuel temperatures of 1800{degree}C, however, fission product release (cesium) from individual particles was significant and there was large particles-to-particle variation. At 1800{degree}C, individual particle release (cesium) was on average ten times the Kernforschungsanlage-measured integral spherical fuel element release value. Particle release data from the sphere tested at 1800{degree}C indicate that there may be two distinct modes of failure at fuel temperatures of 1800{degree}C and above. 5 refs., 9 figs., 9 tabs.
Date: March 1, 1990
Creator: Baldwin, C.A. & Kania, M.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

THE RECOVERY OF FISSION PRODUCT RARE EARTH SULFATES FROM PUREX 1WW

Description: Cerium- and 144 promethium-147, accompanied by rare earths resulting from fission or decay can be removed from Purex 1WW in>90% yield as an insoluble, crystalline sodium-rare earth double sulfate. Precipitation is initiated by a one-to-three hour equilibration at 90 deg C and centrifugation at 90 deg C to take advantage of the lower solubility of the double sulfate salt at a higher temperature. The sulfate concentration should be one molar and the solution pH at the time of precipitation should be 0.5 to 1.5. The addition of tartrate ion to complex the iron allows the use of a higher pH and sulfate concentration, gives a more complete separation from iron, and a quantitative recovery of the rare earths. The double sulfate precipitate can be dissolved in dilute nitric acid or converted to the carbonate and then dissolved to yield a solution for further processing. The double sulfate precipitation of the rare earths, with tartrate added, gives a good separation from impurities. One-cycle decontamination factors of 150 for Zr-Nb and 1100 for Ru-Rh have been achieved in laboratory tests. Tests in the Purex head-end equipment with up to twomegacurie batches of cerium have corroborated the laboratory results. Decontamination factors of 70 for iron, 10 for zirconium, 20 for niobium and 25 for ruthenium have been obtained. It was found wise to limit the batch size because decay heat leads to partial calcination in the centrifuge and to difficulty in redissolution. (auth)
Date: May 10, 1961
Creator: Wheelwright, E.J. & Swift, W.H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

IRRADIATION EFFECTS ON UC$sub 2$ DISPERSED IN GRAPHITE. Final Report

Description: Postirradiation examinations were performed on four fuel cylinders containing UC/sub 2/ dispersed in graphite and sealed in low permeability graphite cans irradiated in the MTR. In all but one experiment, a flowing stream of He passed over the outside of the graphite cans. The estimated fuel cylinder central temperatures ranged from 2500 to 3500 deg F, and the maximum U/sup 235/ burnup was 21%. The most important radiation effects were fuel dimensional change, fission product and U migration, and microscopic structural changes. The fuel cylinders, in general, decreased in diameter and increased in lergth. The decrease in diameter was a function of fuel burnup. A 4.5% decrease in diameter was observed at a burnup of 6 x 10/sup 19/ of the fuel cylinders into the graphite cans. Fission products likewise were found throughout the walls of the graphite can, with steep concentration gradients near the inside surface of the can wall. The istopes Zr/sup 95/ and Ce/sup 144/ were more abundant near the outside wall surface than at midwall. There was no corresponding rise in the concentrations of Sr/sup 90/ and Cs/sup 137/ near the outside surface. Microstructural changes were observed in the UC/sub 2/ particles which were a function of burnup. At around 4 at.% U burnup the particles became porous and began to lose their angular shape. At 21 at.% U burnup, the particles had spheroidized and had begun to diffuse into the surrounding graphite matrix. (auth)
Date: January 1, 1963
Creator: Morgan, J.G. & Osborne, M.F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department