Sample records for gas fission products

  1. Transport of fission products with a helium gas-jet at TRIGA-SPEC

    NASA Astrophysics Data System (ADS)

    Eibach, M.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Herfurth, F.; Geppert, C.; Ketelaer, J.; Ketter, J.; Krmer, J.; Krieger, A.; Knuth, K.; Nagy, Sz.; Nrtershuser, W.; Smorra, C.

    2010-02-01

    A helium gas-jet system for the transport of fission products from the research reactor TRIGA Mainz has been developed, characterized and tested within the TRIGA-SPEC experiment. For the first time at TRIGA Mainz carbon aerosol particles have been used for the transport of radionuclides from a target chamber with high efficiency. The radionuclides have been identified by means of ?-spectroscopy. Transport time, efficiency as well as the absolute number of transported radionuclides for several species have been determined. The design and the characterization of the gas-jet system are described and discussed.

  2. A Simulation Experiment of the Results of Fission Product Gas Release due to Fuel Cladding Failure using Water Circulation System

    Microsoft Academic Search

    Shunsuke KONDO; Kimihide MIYAGUCHI; Shigehiro AN; Akira OYAMA

    1970-01-01

    To examine the effects of fission product gas release into the coolant channel due to fuel cladding failure, two preliminary simulation experiments were undertaken using an electrical heater pin set in a system for water circulation. The first experiment represented a continuous release of gas from a small hole in the cladding, and the second a sudden burst of gas

  3. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  4. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John (Jack) K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/Bs) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  5. A FISSION GAS MONITOR FOR GAS COOLED OR OFF-GAS EXPERIMENTAL SYSTEMS

    Microsoft Academic Search

    J. F. Sommers; I. J. Wells

    1959-01-01

    The fission gas monitor is a very sensitive instrument for the detection ; of fresh fission products in gas cooled or off-gas systems. It is quite ; versatile so that it can be used on a wide variety of experiments either as a ; quick, positive, and sensitive detector of fission products or as a quantitative ; measuring device for

  6. Fission product induced swelling of U-Mo alloy fuel

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Hofman, G. L.

    2011-12-01

    Fuel swelling of U-Mo alloy was modeled using the measured data from samples irradiated up to a fission density of 7 10 27 fissions/m 3 at temperatures below 250 C. The overall fuel swelling was measured from U-Mo foils with as-fabricated thickness of 250 ?m. Volume fractions occupied by fission gas bubbles were measured and fuel swelling caused by the fission gas bubbles was quantified. The portion of fuel swelling by solid fission products including solid and liquid fission products as well as fission gas atoms not enclosed in the fission gas bubbles is estimated by subtracting the portion of fuel swelling by gas bubbles from the overall fuel swelling. Empirical correlations for overall fuel swelling, swelling by gas bubbles, and swelling by solid fission products were obtained in terms of fission density.

  7. Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

    NASA Astrophysics Data System (ADS)

    Johnson, L.; Gnther-Leopold, I.; Kobler Waldis, J.; Linder, H. P.; Low, J.; Cui, D.; Ekeroth, E.; Spahiu, K.; Evins, L. Z.

    2012-01-01

    Studies of the rapid aqueous release of fission products from UO 2 and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of 137Cs and 129I for a number of fuels irradiated to burn-ups of 50-75 MWd/kgU. In addition a method for analysis of leaching solutions for 79Se was developed. The results of the studies show that the fractional release of 137Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional 129I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association 137Cs and 129I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for 79Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that 79Se is not preferentially leached from spent fuel.

  8. Production of fissioning uranium plasma to approximate gas-core reactor conditions

    NASA Technical Reports Server (NTRS)

    Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

    1974-01-01

    The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

  9. Mass spectrometry studies of fission product behavior: 2, Gas phase species

    SciTech Connect

    Blackburn, P.E.; Johnson, C.E.

    1987-01-01

    Revaporization of fission products from reactor system surfaces has become a complicating factor in source term definition. Critical to this phenomena is understanding the nature and behavior of the vapor phase species. This study characterizes the stability of the CsI . CsOH vapor phase complex. Vapor pressures were measured with a mass spectrometer. Thermodynamic data were obtained for CsOH(g), Cs/sub 2/(OH)/sub 2/(g), CsI(g), Cs/sub 2/I/sub 2/(g) and CsI . CsOH(g). Activity coefficients were derived for the CsI-CsOH system. The relative ionization cross section of CsOH is about ten times the cross section of CsI(g). CsI . CsOH fragments to Cs/sub 2/OH/sup +/ and an iodine atom. 17 refs., 4 figs., 6 tabs.

  10. Mobility of fission gas bubbles

    Microsoft Academic Search

    F. A. Nichols; C. Ronchi

    1985-01-01

    The importance of bubble migration in fuel swelling and fission-product release remains a controversial topic in spite of a great deal of research. For steady-state analyses some authors ignore bubble motion totally, whereas others employ mobilities (based on out-of-pile measurements) which are far below the theoretical diffusion-control predictions. Under transient conditions some continue to employ zero or low bubble mobilities,

  11. Fission gas release in LWR fuel measured during nuclear operation

    Microsoft Academic Search

    A. D. Appelhans; E. Skattum; D. J. Osetek

    1980-01-01

    A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods

  12. Fission products of transuranium elements in space

    NASA Astrophysics Data System (ADS)

    Shukoliukov, Iu. A.; Dang, V. M.

    The current status of the problem of transuranium elements in space is reviewed. Particular attention is given to Pu-244 fission products in eucrite achondrites; xenon as a fission product in chondrites; the isotopic composition of Xe in meteorites and the evolution of the earth's atmosphere; anomalies of the isotopic composition of uranium in meteorites; the contribution of fission products to the abundance of other elements in space; fission products in lunar matter; and cosmochronology according to fission products.

  13. Payload dose rate from direct beam radiation and exhaust gas fission products. [for nuclear engine for rocket vehicles

    NASA Technical Reports Server (NTRS)

    Capo, M. A.; Mickle, R.

    1975-01-01

    A study was made to determine the dose rate at the payload position in the NERVA System (1) due to direct beam radiation and (2) due to the possible effect of fission products contained in the exhaust gases for various amounts of hydrogen propellant in the tank. Results indicate that the gamma radiation is more significant than the neutron flux. Under different assumptions the gamma contribution from the exhaust gases was 10 to 25 percent of total gamma flux.

  14. FISSION-PRODUCT CAPTURE CROSS SECTIONS

    Microsoft Academic Search

    J. D. Garrison; B. W. Roos

    1962-01-01

    Experimental measurements of fission product capture cross sections and ; statistical estimates of capture cross sections for energies at which no ; measurements have been made yielded a set of group cross sections for primary and ; secondary fission products covering the complete range of energies of interest ; for reactor calculations. Capture cross sections and fission product yield ;

  15. Fission products of transuranium elements in space

    Microsoft Academic Search

    Iu. A. Shukoliukov; V. M. Dang

    1984-01-01

    The current status of the problem of transuranium elements in space is reviewed. Particular attention is given to Pu-244 fission products in eucrite achondrites; xenon as a fission product in chondrites; the isotopic composition of Xe in meteorites and the evolution of the earth's atmosphere; anomalies of the isotopic composition of uranium in meteorites; the contribution of fission products to

  16. The SPIDER fission fragment spectrometer for fission product yield measurements

    NASA Astrophysics Data System (ADS)

    Meierbachtol, K.; Tovesson, F.; Shields, D.; Arnold, C.; Blakeley, R.; Bredeweg, T.; Devlin, M.; Hecht, A. A.; Heffern, L. E.; Jorgenson, J.; Laptev, A.; Mader, D.; O`Donnell, J. M.; Sierk, A.; White, M.

    2015-07-01

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) has been developed for measuring mass yield distributions of fission products from spontaneous and neutron-induced fission. The 2E-2v method of measuring the kinetic energy (E) and velocity (v) of both outgoing fission products has been utilized, with the goal of measuring the mass of the fission products with an average resolution of 1 atomic mass unit (amu). The SPIDER instrument, consisting of detector components for time-of-flight, trajectory, and energy measurements, has been assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Individual measurement resolutions were met for time-of-flight (250 ps FWHM), spacial resolution (2 mm FHWM), and energy (92 keV FWHM for 8.376 MeV). Mass yield results measured from 252Cf spontaneous fission products are reported from an E-v measurement.

  17. New instrumental method for determining noble fission gas retained in irradiated nuclear fuels

    Microsoft Academic Search

    1981-01-01

    The measurement of fission products generated in nuclear fuel is necessary for the complete characterization of the irradiated fuel. The gaseous fission products, xenon and krypton, are of particular importance. A new method has been developed for the measurement of the fission gas retained in nuclear fuel. The method involves extraction of xenon and krypton by melting the fuel in

  18. Fill gas and fission gas behavior in IFA429

    Microsoft Academic Search

    D. E. Owen; R. W. Miller

    1978-01-01

    The results to date (startup to approximately 23,000 MWd\\/t) of the fill gas absorption and fission gas release experiment IFA-429 are reviewed. The influence of fuel density, diametral gap, and UO grain size on the internal pressure of pressurized fuel rods is discussed. Postirradiation examination measurements of helium fill gas absorption by the fuel, fuel densification, and fission gas release

  19. Fission-Product Yields following Fast Fission of ^238U.^*

    NASA Astrophysics Data System (ADS)

    Campbell, J. M.; Couchell, G. P.; Li, S.; Nguyen, H. V.; Pullen, D. J.; Seabury, E. H.; Schier, W. A.; Tipnis, S. V.; England, T. R.

    1996-10-01

    High-resolution gamma-ray spectra from fast fission of ^238U have been measured at 13 delay-time intervals ranging from 0.3s to 5,000s after fission. The spectra were measured using a high-purity germanium detector enclosed in a NaI(Tl) Compton suppression annulus. The rapid transfer of fission products from the fission chamber to a low-background counting room by means of a helium-jet/tape transport system leads to a marked reduction in background and allows measurement of spectra at short delay times. Beta-gamma coincidence leads to a further reduction in background. Cumulative and independent yields of individual fission products are calculated from the relative line intensities extracted from the aggregate spectra, and are compared to ENDF/B-VI yield values. Supported in part by the U.S. Department of Energy

  20. Fission gas retention in irradiated metallic fuel

    Microsoft Academic Search

    G. R. Fenske; E. E. Gruber; J. M. Kramer

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by

  1. RELEASE OF FISSION GASES FROM IRRADIATED URANIUM DIOXIDE. PART I. APPARATUS FOR THE MEASUREMENT OF FISSION GAS RELEASE FROM FUEL MATERIALS DURING PILE IRRADIATION

    Microsoft Academic Search

    J. M. Markowitz; R. C. Koch; J. A. Roll

    1957-01-01

    An apparatus for the continuous measurement of the rate of escape of ; short-lived inert gas fission products from fuel materials during pile ; irradiation is described. Helium carrier gas is swept over a bare fuel element ; specimen placed in an in-pile furnace. Escaping fission products are carried ; into an out-of-pile gas handling system, where they are removed

  2. Nondestructive fission gas release measurement and analysis

    Microsoft Academic Search

    P. M. OLeary; D. R. Packard

    1993-01-01

    Siemens Power Corporation (SPC) has performed reactor poolside gamma scanning measurements of fuel rods for fission gas release (FGR) detection for more than 10 yr. The measurement system has been previously described. Over the years, the data acquisition system, the method of spectrum analysis, and the means of reducing spectrum interference have been significantly improved. A personal computer (PC)-based multichannel

  3. DIFFUSION OF FISSION GAS IN URANIUM

    Microsoft Academic Search

    1963-01-01

    The diffusion coefficient and activation energy for diffusion of xenon-; 133 in uranium metal have been determined for a number of varying sample ; conditions, with metal of hlgh purity and commercial purity. Thin foil samples ; were neutron irradiated, in groups, under conditions such as to produce optimum ; uniformity of fission product concentration throughout the samples. The samples

  4. Multiscale development of a fission gas thermal conductivity model: Coupling atomic, meso and continuum level simulations

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Millett, Paul C.; Nerikar, Pankaj; Du, Shiyu; Andersson, David; Stanek, Christopher R.; Gaston, Derek; Andrs, David; Williamson, Richard

    2013-09-01

    Fission gas production and evolution significantly impact the fuel performance, causing swelling, a reduction in the thermal conductivity and fission gas release. However, typical empirical models of fuel properties treat each of these effects separately and uncoupled. Here, we couple a fission gas release model to a model of the impact of fission gas on the fuel thermal conductivity. To quantify the specific impact of grain boundary (GB) bubbles on the thermal conductivity, we use atomistic and mesoscale simulations. Atomistic molecular dynamic simulations were employed to determine the GB thermal resistance. These values were then used in mesoscale heat conduction simulations to develop a mechanistic expression for the effective GB thermal resistance of a GB containing gas bubbles, as a function of the percentage of the GB covered by fission gas. The coupled fission gas release and thermal conductivity model was implemented in Idaho National Laboratory's BISON fuel performance code to model the behavior of a 10-pellet LWR fuel rodlet, showing how the fission gas impacts the UO2 thermal conductivity. Furthermore, additional BISON simulations were conducted to demonstrate the impact of average grain size on both the fuel thermal conductivity and the fission gas release.

  5. Reactor power history from fission product signatures

    E-print Network

    Sweeney, David J.

    2009-05-15

    The purpose of this research was to identify fission product signatures that could be used to uniquely identify a specific spent fuel assembly in order to improve international safeguards. This capability would help prevent and deter potential...

  6. Systematics of Fission-Product Yields

    SciTech Connect

    A.C. Wahl

    2002-05-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

  7. Fission Gas Release from Uranium Heated in Carbon Dioxide

    Microsoft Academic Search

    Kiyoaki TAKETANI; Katsuichi IKAWA

    1965-01-01

    An experiment to predict the amount of fission gas release from U heated in CO2 was made in connection with the safely evaluation of the Tokai Atomic Power Reactor. Fission gas release from slightly irradiated U pieces was measured as a function of percent oxidation, and it was found that the percent release was proportional to the percent oxidation. The

  8. Enrichment, separation, and gas-chromatographic and mass-spectrometric analyses of fission products from irradiated or heated fats, oils, and test substances

    Microsoft Academic Search

    Beck

    1973-01-01

    From international colloquium: the identification of irradiated ; foodstuffs; Karlsrahe, Germany (24 Oct 1973). Tripalmitate, tristearate, ; trioleate, oleic acid methyl ester, linoleic acid methyl ester, lauric acid, ; lard, coconut butter, sunflower oil, and olive oil were irradiated at 0.5-6 ; Mrad,or heated up to 174 deg C for 24 hr. The fission products were fractionally ; distilled with

  9. AMS measurements of fission products at CIAE

    NASA Astrophysics Data System (ADS)

    Shen, Hongtao; Jiang, Shan; He, Ming; Dong, Kejun; Ouyang, Yinggen; Li, Zhenyu; Guan, Yongjing; Yin, Xinyi; Peng, Bo; Zhou, Duo; Yuan, Jian; Wu, Shaoyong

    2013-01-01

    Fission products are present in special nuclear materials as contaminants remaining from isotope separation or reprocessing, or through ingrowth due to spontaneous and neutron induced fission. The long half-lived fission products (LLFPs) are among the most dangerous radionuclides to the environment. Ultra-high-sensitivity measurement of LLFPs in rocks or soil samples from the fission environment would provide very important information for nuclear safety inspection. The Beijing HI-13-AMS facility with a high terminal voltage of 13 MV is suitable for measuring LLFPs, especially for heavy fission products such as 79Se, 93Zr, 99Tc, 107Pd, 121mSn, 126Sn, 129I and 151Sm. In this paper some new methods developed for AMS measurement of 79Se, 93Zr, 99Tc, 121mSn, 126Sn, 129I and 151Sm are presented. Major features of these methods will be introduced, including the preparation of samples, the selection of target material and the molecular ions extracted from the material in the ion source, as well as the identification and detection of the nuclides to be determined.

  10. Nondestructive fission gas release measurement and analysis

    SciTech Connect

    O'Leary, P.M.; Packard, D.R. (Siemens Nuclear Power Corp., Richland, WA (United States))

    1993-01-01

    Siemens Power Corporation (SPC) has performed reactor poolside gamma scanning measurements of fuel rods for fission gas release (FGR) detection for more than 10 yr. The measurement system has been previously described. Over the years, the data acquisition system, the method of spectrum analysis, and the means of reducing spectrum interference have been significantly improved. A personal computer (PC)-based multichannel analyzer (MCA) package is used to collect, display, and store high-resolution gamma-ray spectra measured in the fuel rod plenum. A PC spread sheet is used to fit the measured spectra and compute sample count rates after Compton background subtraction. A Zircaloy plenum spacer is often used to reduce positron annihilation interference that can arise from the INCONEL[sup [reg sign

  11. PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA

    E-print Network

    was slow. PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA In relation to "fallout" from nuclear -bomb the penetration of radioactive strontium, cesium, and ruthenium common products of nuclear fission, through

  12. Rapid separation of fresh fission products (draft)

    SciTech Connect

    Dry, D. E. (Donald E.); Bauer, E. (Eve); Petersen, L. A. (Lisa A.)

    2003-01-01

    The fission of highly eruiched uranium by thermal neutrons creates dozens of isotopic products. The Isotope and Nuclear Chemistry Group participates in programs that involve analysis of 'fiesh' fission products by beta counting following radiochemical separations. This is a laborious and time-consuming process that can take several days to generate results. Gamma spectroscopy can provide a more immediate path to isolopic activities, however short-lived, high-yield isotopes can swamp a gamma spectrum, making difficult the identification and quantification of isotopes on the wings and valley of the fission yield curve. The gamma spectrum of a sample of newly produced fission products is dominated by the many emissions of a very few high-yield isotopes. Specilkally, {sup 132}Te (3.2 d), its daughter, {sup 132}I(2 .28 h), {sup 140}Ba (12.75 d), and its daughter {sup 140}La (1.68 d) emit at least 18 gamma rays above 100 keV that are greater than 5% abundance. Additionally, the 1596 keV emission fiom I4'La imposes a Compton background that hinders the detection of isotopes that are neither subject to matrix dependent fractionation nor gaseous or volatile recursors. Some of these isotopes of interest are {sup 111}Ag, {sup 115}Cd, and the rare earths, {sup 153}Sm, {sup 154}Eu, {sup 156}Eu, and {sup 160}Tb. C-INC has performed an HEU irradiation and also 'cold' carrier analyses by ICP-AES to determine methods for rapid and reliable separations that may be used to detect and quantify low-yield fission products by gamma spectroscopy. Results and progress will be presented.

  13. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J. (Los Alamos National Laboratory, MS E550, Los Alamos, New Mexico (USA))

    1991-01-10

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere.

  14. Modeling fission-gas behavior in metallic fuels

    Microsoft Academic Search

    EUGENE E. GRUBER; JOHN M. KRAMER

    1987-01-01

    The FRAS3 code, developed to model transient fission-gas behavior in oxide fuel, has been modified for application to metallic fuel. The code has been applied to analyze a series of experiments on irradiated uranium fuel, in which fission-gas bubble distributions were measured following isothermal anneals under various pressures. Comparison of the calculated bubble-size distributions to those measured indicates that grain-boundary

  15. Mechanisms of in-pile fission gas release from UO

    Microsoft Academic Search

    K. Shiba; M. Handa; S. Yamagishi; T. Fukuda; Y. Takahashi; T. Tanifuji; S. Omori; A. Kondo

    1973-01-01

    In-pile release mechanisms of fission gas from UO at low ; temperatures were studied. The release of ¹³³Xe, ¹³⁵Xe, ¹³⁸Se, ; 2\\/ pellet was measured at temperatures ranging from 250 to 930 deg C using a ; graphite specimen holder. The release from the holder, in which a fraction of ; fission gas recoil-implanted, was subtracted to obtain the net

  16. Mass distribution of fission products following photofission of uranium-238

    Microsoft Academic Search

    D. Swindle; R. Wright; K. Takahashi; W. H. Rivera; J. L. Meason

    1973-01-01

    The mass-yield distribution of fission products following photofission ; of ²³⁸U using bremsstrahlung energies of 22, 24, and 28 MeV were measured ; by radiochemically isolating the fission products belonging to 24 mass chains. ; The absolute activities of these nuclides were determined by BETA - and gamma ; counting techniques, and the cumulative fission yields were calculated relative ;

  17. On-site ? -ray spectroscopic measurements of fission gas release in irradiated nuclear fuel

    Microsoft Academic Search

    I. Matsson; B. Grapengiesser; B. Andersson

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514keV ?-radiation from the gaseous fission product 85Kr captured in the fuel rod plenum volume. A submergible

  18. Quality and availability of fission-products

    Microsoft Academic Search

    Rasmussen

    1965-01-01

    The information contained in this report was copied from carts used for discussions at Richland on May 4--6, 1965. Representatives of the Martin Company, US Rubber company, General Electric -- HAPO, and AEC-RL00 were present. The data represent the best current information on the quantity, quality, and availability of each fission product considered, i.e., strontium-90, cesium-137, promethium-147, and cerium-144.

  19. Time dependent particle emission from fission products

    SciTech Connect

    Holloway, Shannon T [Los Alamos National Laboratory; Kawano, Toshihiko [Los Alamos National Laboratory; Moller, Peter [Los Alamos National Laboratory

    2010-01-01

    Decay heating following nuclear fission is an important factor in the design of nuclear facilities; impacting a variety of aspects ranging from cooling requirements to shielding design. Calculations of decay heat, often assumed to be a simple product of activity and average decay product energy, are complicated by the so called 'pandemonium effect'. Elucidated in the 1970's this complication arises from beta-decays feeding high-energy nuclear levels; redistributing the available energy between betas and gammas. Increased interest in improving the theoretical predictions of decay probabilities has been, in part, motivated by the recent experimental effort utilizing the Total Absorption Gamma-ray Spectrometer (TAGS) to determine individual beta-decay transition probabilities to individual nuclear levels. Accurate predictions of decay heating require a detailed understanding of these transition probabilities, accurate representation of particle decays as well as reliable predictions of temporal inventories from fissioning systems. We will discuss a recent LANL effort to provide a time dependent study of particle emission from fission products through a combination of Quasiparticle Random Phase Approximation (QRPA) predictions of beta-decay probabilities, statistical Hauser-Feshbach techniques to obtain particle and gamma-ray emissions in statistical Hauser-Feshbach and the nuclear inventory code, CINDER.

  20. SPIDER Progress Towards High Resolution Correlated Fission Product Data

    NASA Astrophysics Data System (ADS)

    Shields, Dan; Meierbachtol, Krista; Tovesson, Fredrik; Arnold, Charles; Blackeley, Rick; Bredeweg, Todd; Devlin, Matt; Hecht, Adam; Jandel, Marian; Jorgenson, Justin; Nelson, Ron; White, Morgan; Spider Team

    2014-09-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (?), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) is under development with the goal of obtaining high-resolution, high-efficiency, correlated fission product data needed for many applications including the modeling of next generation nuclear reactors, stockpile stewardship, and the fundamental understanding of the fission process. SPIDER simultaneously measures velocity and energy of both fission products to calculate fission product yields (FPYs), neutron multiplicity (?), and total kinetic energy (TKE). A detailed description of the prototype SPIDER detector components will be presented. Characterization measurements with alpha and spontaneous fission sources will also be discussed. LA-UR-14-24875. This work is in part supported by LANL Laboratory Directed Research and Development Projects 20110037DR and 20120077DR.

  1. Fuel and fission gas response to simulated thermal transients: experimental results and correlation with fission gas release and swelling model. [LMFBR

    Microsoft Academic Search

    Bandyopadhyay

    1978-01-01

    To evaluate the role of fission gas in hypothetical core disruptive accidents, experimental and analytical information describing the fission gas behavior in rapid temperature transients is urgently needed. A direct-electrical-heating apparatus was used to obtain information on the fission gas behavior and the response of mixed-oxide fuel elements to simulated thermal transient conditions. The experimental results indicate that fission gas

  2. Experimental investigation of coolability degradation by fission gas release into flowing sodium in a fuel pin bundle

    Microsoft Academic Search

    K. Haga; Y. Kikuchi

    1985-01-01

    A series of experiments was performed to assess the thermal effect of a burst-type fission gas release from fuel pins. Simulated fission product gas was injected continuously and transiently from the central pin of a 37-pin bundle. The opposite pin surface impinged on by the released gas showed an extreme temperature rise under high coolant-flow conditions. Comparison of measured temperature

  3. Fission-gas-release rates from irradiated uranium nitride specimens

    NASA Technical Reports Server (NTRS)

    Weinstein, M. B.; Kirchgessner, T. A.; Tambling, T. N.

    1973-01-01

    Fission-gas-release rates from two 93 percent dense UN specimens were measured using a sweep gas facility. Specimen burnup rates averaged .0045 and .0032 percent/hr, and the specimen temperatures ranged from 425 to 1323 K and from 552 to 1502 K, respectively. Burnups up to 7.8 percent were achieved. Fission-gas-release rates first decreased then increased with burnup. Extensive interconnected intergranular porosity formed in the specimen operated at over 1500 K. Release rate variation with both burnup and temperature agreed with previous irradiation test results.

  4. APPARATUS FOR THE STUDY OF FISSION-GAS RELEASE FROM FUELS DURING POSTIRRADIATION HEATING AT TEMPERATURES UP TO 1600 C

    Microsoft Academic Search

    R. H. Barnes; D. N. Sunderman

    1960-01-01

    An apparatus to study rare-gas fission-product release from nuclear fuel ; materials during postirradiation heating was developed. Xenon and krypton ; fission gases escaping from a small specimen during heating at constant ; temperature are measured using a continuous radioactivity monitor and charcoal ; adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg ; C. Helium carrier

  5. New instrumental method for determining noble fission gas retained in irradiated nuclear fuels

    SciTech Connect

    Baldwin, D L

    1981-01-01

    The measurement of fission products generated in nuclear fuel is necessary for the complete characterization of the irradiated fuel. The gaseous fission products, xenon and krypton, are of particular importance. A new method has been developed for the measurement of the fission gas retained in nuclear fuel. The method involves extraction of xenon and krypton by melting the fuel in a commercially available furnace. Several factors influence the complete fusion of the fuel and release of the noble gases. Development work aimed at identifying and understanding these factors is discussed. The gases are purified after release from the fuel and collected on cryogenically-cooled activated charcoal. The gases are subsequently released from the charcoal trap and measured by gas chromatography. Column requirements and optimum operating conditions are discussed. Various modifications to the furnace are necessary for reliable performance within the high radiation environment. Other radiological problems are identified and their solutions discussed.

  6. ENDF/B fission-product data

    SciTech Connect

    Rose, P.F.; Burrows, T.W.

    1981-01-01

    In recognition of the pressing need for nuclear decay data in a variety of reactor-related applications, the scope of the Evaluated Nuclear Data Files (ENDF/B) was expanded to include such information. The initial reason for this expansion was to provide reliable data and a common data base to be used in computer codes developed to carry out summation calculations of the decay-heat source term in reactor cores. The ENDF/B-V fission product file was released in July 1980 and contains data for 877 nuclides. Forty-two of these are found in greater detail on the General Purpose file. Seven hundred and fifty isotopes contain decay data. One hundred and ninety-six contain cross section data for total, elastic, inelastic, and angular distributions in the energy range 10/sup -5/ eV to 20 MeV. This special fission product file is contained on 6 magnetic tapes, and is available from the National Nuclear Data Center (NNDC) at BNL. (WHK)

  7. URANIUM235 FISSION-PRODUCT PRODUCTION AS A FUNCTION OF THE THERMAL NEUTRON FLUX, IRRADIATION TIME, AND DECAY TIME. II. SUMMATIONS OF INDIVIDUAL CHAINS, ELEMENTS, AND THE RARE-GAS AND RARE-EARTH GROUPS. VOLUMES 1 AND 2

    Microsoft Academic Search

    J. O. Blomeke; M. F. Todd

    1958-01-01

    These two volumes were issued separately, but are cataloged as a unit. ; The following properties, per initial atom of U²³⁵, are tabulated for each ; fission-product chain with mass number 72 to 161: activity, gamma power, total ; power, poisoning, and gamma disintegrations per second with energies up to 1.70 ; Mev. (M.H.R.);

  8. Fission product behavior in the Molten Salt Reactor Experiment

    Microsoft Academic Search

    E. L. Compere; S. S. Kirslis; E. G. Bohlmann; F. F. Blankenship; W. R. Grimes

    1975-01-01

    Essentially all the fission product data for numerous and varied samples ; taken during operation of the Molten Salt Reactor Experiment or as part of the ; examination of specimens removed after particular phases of operation are ; reported, together with the appropriate inventory or other basis of comparison, ; and relevant reactor parameters and conditions. Fission product behavior fell

  9. Chemical factors affecting fission product transport in severe LMFBR accidents

    SciTech Connect

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

  10. The growth of fission gas bubbles in irradiated uranium dioxide

    Microsoft Academic Search

    R. M. Cornell

    1969-01-01

    The growth of fission gas bubbles from supersaturated solution in irradiated uranium dioxide has been studied by electron microscopy under isothermal annealing conditions between 1300 and 1500C. Measurements of the kinetics of bubble growth have enabled the diffusion coefficients of atomic xenon and krypton in irradiated uranium dioxide to be determined. The diffusion coefficients obtained may be expressed by the

  11. (Fuel, fission product, and graphite technology)

    SciTech Connect

    Stansfield, O.M.

    1990-07-25

    Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

  12. Analysis of Fission Products on the AGR-1 Capsule Components

    SciTech Connect

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.210 2 (Capsule 3) to 3.810 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  13. Fission gas and void volume measuring system

    Microsoft Academic Search

    1974-01-01

    From joint meeting of the American Nuclear Society and the Atomic ; lndustrial Forum and Nuclear Energy Exhibition; San Francisco, California, USA ; (11 Nov 1973). A system which provides a measurement of the amount of gas and ; free void volume contained in encapsulated fuel elements has been in use at LASL ; for over a year. A gas

  14. Search for production of superheavy elements via fusion after instantaneous fission in the reaction 238 U+ 208 Pb

    Microsoft Academic Search

    T. Lund; D. Hirdes; H. Jungclas; D. Molzahn; P. Vater; R. Brandt; P. Lemmertz; R. Fa; H. Wollnik; H. Gggeler

    1981-01-01

    Superheavy elements (SHE) might be formed via a reaction mechanism called fusion after instantaneous fission, which is supposed to occur during the collision of a deformed very heavy nucleus with a spherical one. We bombarded natural uranium targets with lead ions and searched for alpha-emitting and spontaneously fissioning reaction products. Different techniques were used: a rotating wheel, a gas jet

  15. Measurement of fission product gases in the atmosphere

    Microsoft Academic Search

    W. R. Schell; M. J. Tobin; D. J. Marsan; C. W. Schell; J. Vives-Batlle; S. R. Yoon

    1997-01-01

    The ability to quickly detect and assess the magnitude of releases of fission-produced radioactive material is of significant importance for ongoing operations of any conventional nuclear power plant or other activities with a potential for fission product release. In most instances, the control limits for the release of airborne radioactivity are low enough to preclude direct air sampling as a

  16. Preliminary results utilizing high-energy fission product ?-rays to detect fissionable material in cargo

    NASA Astrophysics Data System (ADS)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Church, J. A.; Descalle, M. A.; Gosnell, T. B.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Mauger, G. J.; Moore, T. L.; Norman, E. B.; Pohl, B. A.; Pruet, J. A.; Petersen, D. C.; Walling, R. S.; Weirup, D. L.; Prussin, S. G.; McDowell, M.

    2005-12-01

    A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their ?-delayed neutron emission or ?-delayed high-energy ? radiation between beam pulses provide the detection signature. Fission product ?-delayed ?-rays above 3 MeV are nearly 10 times more abundant than ?-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified.

  17. (COMEDIE program review and fission product transport in MHTGR reactor)

    SciTech Connect

    Stansfield, O.M.

    1990-03-15

    The subcontract between Martin Marietta Energy Systems, Inc., and the CEA provides for the refurbishment of the high pressure COMEDIE test loop in the SILOE reactor and a series of experiments to characterize fission product lift-off from MHTGR heat exchanger surfaces under several depressurization accident scenarios. The data will contribute to the validation of models and codes used to predict fission product transport in the MHTGR. In the meeting at CEA headquarters in Paris the program schedule and preparation for the DCAA and Quality Assurance audits were discussed. Long-range interest in expanded participation in the gas-cooled reactor technology Umbrella Agreement was also expressed by the CEA. At the CENG, in Grenoble, technical details on the loop design, fabrication components, development of test procedures, and preparation for the DOE quality assurance (QA) audit in May were discussed. After significant delays in CY 1989 it appears that good progress is being made in CY 1990 and the first major test will be initiated by December. An extensive list of agreements and commitments was generated to facilitate the coordination and planning of future work. 2 figs., 2 tabs.

  18. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  19. Intermediate model on intragranular fission-gas behavior during steady-state irradiation of LMFBR uranium-carbide nuclear fuel

    Microsoft Academic Search

    1980-01-01

    A reliable and computationally efficient physically based model is developed to study the phenomena which regulate intragranular fission gas behavior in LMFBR uranium carbide fuel under operational conditions. Fission gas atoms diffuse in the grain matrix and continuously precipitate into immobile clusters-fission gas bubble embryos-, by agglomeration of two gas atoms. Embryos may survive and grow into equal size fission

  20. Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

    SciTech Connect

    J M Harp; P D Demkowicz; S A Ploger

    2012-10-01

    The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INLs Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

  1. Two-step two-stage fission gas release model

    Microsoft Academic Search

    Yong-soo Kim; Chan-bock Lee

    2008-01-01

    A two-step two-stage model is developed in this study on the basis of the recent theoretical model. This model incorporates a two-step burn-up factor in the two-stage diffusion processes in the grain lattice and at the grain boundary during the fission gas release. In-pile data sets available in FRAPCON-3 code are used to validate the model. Results show that the

  2. Effects of Void Migration Process on Fission-Gas Release

    Microsoft Academic Search

    Akira DOI; Kiyoshi INOUE; Noboru HOKKYO; Hiroshi HAYASHI

    1972-01-01

    The release rate of fission-gas from U02 was continuously measured during irradiation in the Hitachi Training Reactor. The U02 specimen was heated electrically in in-core assemblies by tungsten heaters, either arranged axially transversing the specimen (producing radial temperature gradient) or cylindrically outside the specimen (uniform heating). In a case of the axially heated annular U02 pellets with radial temperature gradient,

  3. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect

    Messenger, S. J. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 54-1717, Cambridge, MA 02139 (United States); Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 24-207, Cambridge, MA 02139 (United States); Massie, M. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., NW12-230, Cambridge, MA 02139 (United States)

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. (authors)

  4. Fuel Performance Experience, Analysis and Modeling: Deformations, Fission Gas Release and Pellet-Clad Interaction

    Microsoft Academic Search

    G. Zhou; L. Hallstadius; S. Helmersson; A. R. Massih; D. Schrire; R. Kaellstroem; G. Wikmark; C. Hellwig

    2007-01-01

    Some basic attributes of light water reactor fuel performance, determined by measurements, are evaluated. In particular, data on fuel volume swelling, cladding creep\\/growth, fission product gas release and cladding deformation due to pellet-clad mechanical interaction of rods irradiated in power reactors to rod burnups up to about 70 MWd\\/kgU are presented and appraised. A thermal fuel matrix swelling caused by

  5. Simulating ?-? coincidences of ?-delayed ?-rays from fission product nuclei

    NASA Astrophysics Data System (ADS)

    Padgett, Stephen; Wang, Tzu-Fang

    2015-01-01

    Analyzing radiation from material that has undergone neutron induced fission is important for fields such as nuclear forensics, reactor physics, and nonproliferation monitoring. The ?-ray spectroscopy of fission products is a major part of the characterization of a material's fissile inventory and the energy of incident neutrons inducing fission. Cumulative yields and ?-ray intensities from nuclear databases are inputs into a GEANT4 simulation to create expected ?-ray spectra from irradiated 235U. The simulations include not only isotropically emitted ?-rays but also ?-? cascades from certain fission products, emitted with their appropriate angular correlations. Here ? singles spectra as well as ?-? coincidence spectra are simulated in detectors at both 90 and 180 pairings. The ability of these GEANT4 Monte Carlo simulations to duplicate experimental data is explored in this work. These simulations demonstrate potential in exploiting angular correlations of ?-? cascades in fission product decays to determine isotopic content. Analyzing experimental and simulated ?-? coincidence spectra as opposed to singles spectra should improve the ability to identify fission product nuclei since such spectra are cleaner and contain more resolved peaks when compared to ? singles spectra.

  6. Compilation of fission product yields Vallecitos Nuclear Center

    SciTech Connect

    Rider, B.F.

    1980-01-01

    This document is the ninth in a series of compilations of fission yield data made at Vallecitos Nuclear Center in which fission yield measurements reported in the open literature and calculated charge distributions have been utilized to produce a recommended set of yields for the known fission products. The original data with reference sources, as well as the recommended yields are presented in tabular form for the fissionable nuclides U-235, Pu-239, Pu-241, and U-233 at thermal neutron energies; for U-235, U-238, Pu-239, and Th-232 at fission spectrum energies; and U-235 and U-238 at 14 MeV. In addition, U-233, U-236, Pu-240, Pu-241, Pu-242, Np-237 at fission spectrum energies; U-233, Pu-239, Th-232 at 14 MeV and Cf-252 spontaneous fission are similarly treated. For 1979 U234F, U237F, Pu249H, U234He, U236He, Pu238F, Am241F, Am243F, Np238F, and Cm242F yields were evaluated. In 1980, Th227T, Th229T, Pa231F, Am241T, Am241H, Am242Mt, Cm245T, Cf249T, Cf251T, and Es254T are also evaluated.

  7. FISSION-PRODUCT-RELEASE MEASUREMENT FROM CLAD FUEL SPECIMENS

    Microsoft Academic Search

    M. Kangilaski; A. A. Bauer; F. A. Rough; R. F. Dickerson

    1962-01-01

    Xe¹³³, Xe¹³⁵, Kr\\/sup 85m\\/, Kr⁸⁷, I¹³¹, I¹³³; , and I¹³⁵ release from UO fuel specimens with four types of ; metallic claddings was determined during irradintion in 2200 to 3200 deg F ; flowing helium and oxygen. Fission-gas samples cold trapped from the iurnace ; sweep gas were identified in a gamma analyzer. Informntion on iodine release was ; obtained

  8. ABSORPTION CROSS SECTION OF STABLE, LOW CROSS SECTION FISSION PRODUCTS

    Microsoft Academic Search

    Breslauer

    1959-01-01

    A record of the relatively stable, low cross section fissibn products ; available on ANP nuclear data tape is presented for conventional reactor ; calculations. The record is based on data included in the Geneva conference ; paper by Gordeev and Pupko. The label slag is applied to fission products. ; (J.R.D.);

  9. Comparison of Fission Product Yields and Their Impact

    SciTech Connect

    S. Harrison

    2006-02-01

    This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiological transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.

  10. Fission product release from fuel under severe accident conditions

    SciTech Connect

    Hobbins, R.R.; Petti, D.A.; Hagrman, D.L. (EG and G Idaho, Inc., Idaho Falls (United States))

    1993-03-01

    Recent advances in the understanding of fission product release from fuel under severe accident conditions in light water reactors are reviewed. In addition to the effects of temperature and time at temperature, recent results from in-pile and out-of-pile tests and the accident at Three Mile Island Unit 2 suggest that the effects of fuel morphology such as restructuring of the UO[sub 2] microstructure, fuel liquefaction, molten pool formation, debris bed formation, and the effect of fuel chemistry have important influences on fission product release behavior under severe accident conditions. Consideration of these effects is required for complete models of fission product release during severe light water reactor accidents.

  11. Early results utilizing high-energy fission product (gamma) rays to detect fissionable material in cargo

    SciTech Connect

    Slaughter, D R; Accatino, M R; Bernstein, A; Church, J A; Descalle, M A; Gosnell, T B; Hall, J M; Loshak, A; Manatt, D R; Mauger, G J; McDowell, M; Moore, T M; Norman, E B; Pohl, B A; Pruet, J A; Petersen, D C; Walling, R S; Weirup, D L; Prussin, S G

    2004-09-30

    A concept for detecting the presence of special nuclear material ({sup 235}U or {sup 239}Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their {beta}-delayed neutron emission or {beta}-delayed high-energy {gamma}-radiation between beam pulses provide the detection signature. Fission product {beta}-delayed {gamma}-rays above 3 MeV are nearly ten times more abundant than {beta}-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified. An important goal in the US is the detection of nuclear weapons or special nuclear material (SNM) concealed in intermodal cargo containers. This must be done with high detection probability, low false alarm rates, and without impeding commerce, i.e. about one minute for an inspection. The concept for inspection has been described before and its components are now being evaluated. While normal radiations emitted from plutonium may allow its detection, the majority of {sup 235}U {gamma} ray emission is at 186 keV, is readily attenuated by cargo, and thus not a reliable detection signature for passive detection. Delayed neutron detection following a neutron or photon beam pulse has been used successfully to detect lightly or unshielded SNM targets. While delayed neutrons can be easily distinguished from beam neutrons they have relatively low yield in fission, approximately 0.008 per fission in {sup 239}Pu and 0.017 per fission in {sup 235}U, and are rapidly attenuated in hydrogenous materials making that technique unreliable when challenged by thick hydrogenous cargo overburden. They propose detection of {beta}-delayed high-energy {gamma} radiation as a more robust signature characteristic of SNM.

  12. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    NASA Astrophysics Data System (ADS)

    Selby, H. D.; Mac Innes, M. R.; Barr, D. W.; Keksis, A. L.; Meade, R. A.; Burns, C. J.; Chadwick, M. B.; Wallstrom, T. C.

    2010-12-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99Mo, 95Zr, 137Cs, 140Ba, 141,143Ce, and 147Nd. Modest incident-energy dependence exists for the 147Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by 5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for 99Mo where the present results are about 4%-relative higher for neutrons incident on 239Pu and 235U. Additionally, our results illustrate the importance of representing the incident energy dependence of fission product yields over the fast neutron energy range for high-accuracy work, for example the 147Nd from neutron reactions on plutonium. An upgrade to the ENDF library, for ENDF/B-VII.1, based on these and other data, is described in a companion paper to this work.

  13. Comparison of COMETHE-IIIJ and FCODE-BETA fission gas-release predictions with measurements. Final report. [PWR; BWR

    Microsoft Academic Search

    S. Lee; L. Rayes; E. Rumble; D. Wheeler; A. Woods

    1983-01-01

    This report describes a comparison of the Fission Product Gas Release (FGR) predictability of two LWR fuel rod modeling codes: COMETHE-IIIJ and FCODE-BETA. The comparison is made using 124 well characterized fuel rods with FGR measurements in the EPRI Fuel Performance Data Base.

  14. MARGARET: An Advanced Mechanistic Model of Fission Gas Behavior in Nuclear Fuel

    Microsoft Academic Search

    Laurence NOIROT

    2006-01-01

    CEA (Commissariat l'Energie Atomique) develops several fission gas models with different levels of description (more or less detailed or mechanistic descriptions). After a synthesis of the main phenomena which have to be considered in a fission gas model, the MARGARET model is presented. Up to five cavities populations (bubbles nucleated during irradiation or fabrication pores) are taken into account,

  15. Diffusion Coefficient of Fission Gas in Uranium Dioxide Powder Formed by Carbon Dioxide Oxidation of Uranium

    Microsoft Academic Search

    Kiyoaki TAKETANI; Katsuichi IKAWA

    1965-01-01

    In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO2 oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized

  16. Acoustic Sensor for In-Pile Fuel Rod Fission Gas Release Measurement

    Microsoft Academic Search

    D. Fourmentel; J. F. Villard; J. Y. Ferrandis; F. Augereau; E. Rosenkrantz; M. Dierckx

    2011-01-01

    Innovative in-pile instrumentation is crucial for ad- vanced experimental programs in research reactors. In this field, we developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in Material Testing Reactors (MTR). In order to perform experimental programs related to the study of the fission gas release kinetics,

  17. Acoustic sensor for in-pile fuel rod fission gas release measurement

    Microsoft Academic Search

    D. Fourmentel; J. F. Villard; J. Y. Ferrandis; F. Augereau; E. Rosenkrantz; M. Dierckx

    2009-01-01

    Innovative in-pile instrumentation is crucial for advanced experimental programs in research reactors. this field, we developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French

  18. Ab initio modelling of volatile fission products in uranium mononitride

    NASA Astrophysics Data System (ADS)

    Klipfel, M.; Van Uffelen, P.

    2012-03-01

    Defects and the incorporation of volatile fission products (xenon, krypton, caesium and iodine) in uranium mononitride are investigated using DFT calculations. Various locations for impurities are considered including at a tetrahedral interstitial position, substitution of a host nitrogen or uranium ion and placed in a Schottky defect (UN bivacancy). The incorporation is energetically more favourable for the latter, although the incorporation energies are positive. The preferred position for volatile fission products in UN is at the larger of the vacancies, either a single uranium vacancy or the uranium vacancy of a Schottky defect. The incorporation of a fission product in a bound [1 0 0]-Schottky defect leads to a tetragonal distortion of the supercell. The impurities considered in this work produce very small perturbations of the crystalline matrix of UN. With the exception of impurities at the interstitial site, which perturb the structure into the second coordination sphere, only the displacement of the atoms at the nearest-neighbour positions is significant. Analysis of the charge distribution after incorporation of the fission product reveals a weak charge transfer for the noble gases, while a larger transfer is displayed for caesium and iodine.

  19. Fission properties and production mechanisms for the heaviest known elements

    SciTech Connect

    Hoffman, D.C.

    1981-01-01

    Mass yields of the spontaneous fission of Fm isotopes, Cf isotopes, and /sup 259/Md are discussed. Actinide yields were measured for bombardments of /sup 248/Cm with /sup 16/O, /sup 18/O, /sup 20/Ne, and /sup 22/Ne. A superheavy product might be produced by bombarding /sup 248/Cm with /sup 48/Ca ions. 12 figures. (DLC)

  20. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.

  1. A review of selected aspects of the effect of water vapor on fission gas release from uranium oxycarbide

    Microsoft Academic Search

    1994-01-01

    A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial,

  2. Thermal release of volatile fission products from irradiated nuclear fuel

    SciTech Connect

    Bray, L.A.; Burger, L.L.; Morgan, L.G.; Baldwin, D.L.

    1983-06-01

    An effective procedure for removing /sup 3/H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO/sub 2/ fuel. The release characteristics of /sup 3/H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released about 99% of the /sup 3/H and 20% of both Kr and Xe within a 3 h at 1500/sup 0/C. Similar results were obtained using a carrier gas of He containing 6% H/sub 2/. However, a carrier gas containing only He resulted in the release of approximately 80% of the /sup 3/H and 99% of both Kr and Xe. These results indicate that the release of these volatile fission products from irradiated nuclear fuel is a function of the chemical composition of the gaseous environment. The rate of tritium release increased with increasing temperature (1100 to 1500/sup 0/C) and with the addition of hydrogen to the gas stream. Using crushed UO/sub 2/ fuel without cladding and He as the carrier gas, Kr was completely released at 1500/sup 0/C in 2.5 h. Below 1350/sup 0/C, no Kr-Xe release was observed. Approximately 86% of the /sup 129/I and 95% of the cesium was released from a piece (3.9 g) of UO/sub 2/ fuel at 1500/sup 0/C in He. The zirconium cladding was observed to fracture during heat treatment. A large-scale thermal outgassing system was conceptually designed by the General Atomic Company from an engineering analysis of available experimental data. The direct cost of a 0.5 metric/ton day thermal outgassing system is estimated to be $1,926,000 (1982 dollars), including equipment, installation, instrumentation and controls, piping, and services. The thermal outgassing process was determined to be a technically feasible and cost-competitive process to remove tritium in the head-end portion of a LWR fuel reprocessing plant. Additional laboratory-scale development has been recommended.

  3. The coupled kinetics of grain growth and fission product behavior in nuclear fuel under degraded-core accident conditions

    NASA Astrophysics Data System (ADS)

    Rest, J.

    1985-04-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, and cesium release from (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests (performed at Oak Ridge National Laboratory) and (2) trace-irradiated LWR fuel during severe-fuel-damage (SFD) tests (performed in the PBF reactor in Idaho). A theory of grain boundary sweeping of gas bubbles has been included within the FASTGRASS-VFP formalism. This theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges, and provides a means of determining whether gas bubbles are caught up and moved along by a moving grain boundary or whether the grain boundary is only temporarily retarded by the bubbles and then breaks away. In addition, as FASTGRASS-VFP provides for a mechanistic calculation of ultra- and intergranular fission product behavior, the coupled calculation between fission gas behavior and grain growth is kinetically comprehensive. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. The effect of fuel oxidation by steam on fission product and grain growth behavior is also considered. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted.

  4. GETTER - A MODEL FOR FISSION PRODUCT RELEASE FROM SPHERICAL HTR FUEL ELEMENTS

    Microsoft Academic Search

    Jeetesh Keshaw; Hanno van der Merwe

    Fission product release from spherical fuel spheres under different irradiation and heat-up conditions is one of the key criteria used in High Temperature Reactor (HTR) design. Accurate analyses of fuel performance and fission product behaviour is therefore essential in justifying the safe behaviour of the Pebble Bed Modular Reactor (PBMR). Fuel material parameters and reactor conditions that influence fission product

  5. An assessment of the radiological doses resulting from accidental uranium aerosol releases and fission product releases from a postulated criticality accident at the Oak Ridge Y-12 Plant

    Microsoft Academic Search

    S. E. Fisher; K. E. Lenox

    1995-01-01

    A dose assessment for two separate normalized source terms was conducted for the Oak Ridge Y-12 Plant. The first source term consisted of the noble gas and iodine fission products emanating from a postulated criticality with a magnitude of 10¹⁹ fissions. The second postulated source term was 1 kg of respirable highly enriched uranium. The MELCOR Accident Consequence Code System

  6. Fission-product chemistry in the primary system. [PWR; BWR

    SciTech Connect

    Elrick, R.M.; Sallach, R.A.

    1983-01-01

    Significant retention of fission products in the primary system can occur during severe reactor accidents. Some of these retention processes have been identified, including their reaction rates and reaction products. The reactions investigated include: CsI, CsOH, and tellurium with the structural materials Inconel 600 and 304 stainless steel, tellurium with tin, silver and zircaloy, and boron carbide (B/sub 4/C) in steam and with CsOH and with CsI.

  7. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    SciTech Connect

    G. Pastore; L.P. Swiler; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; L. Luzzi; P. Van Uffelen; R.L. Williamson

    2014-10-01

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code and a recently implemented physics-based model for the coupled fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information from the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior modeling with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  8. Fission-gas release in a spinel-based fuel used for actinide transmutation

    Microsoft Academic Search

    K. Bakker; R. Belvroy; F. A. van den Berg; S. Casalta; R. Conrad; E. A. C. Neeft; R. P. C. Schram; W. Tams

    2001-01-01

    The present paper discusses the fission-gas release as measured from three fuels considered for the transmutation process. The releases were measured in spinel-based fuels from the EFTTRA-T3 irradiation [1,2]. Gas puncturing and EPMA data are discussed for three different inert-matrix concepts. The differences in the fission-gas release for the three spinel-based fuel concepts are most likely connected with the concentrations

  9. Radiation re-solution of fission gas in non-oxide nuclear fuel

    NASA Astrophysics Data System (ADS)

    Matthews, Christopher; Schwen, Daniel; Klein, Andrew C.

    2015-02-01

    Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called 'homogenous' atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

  10. Fuel Performance Experience, Analysis and Modeling: Deformations, Fission Gas Release and Pellet-Clad Interaction

    SciTech Connect

    Zhou, G.; Hallstadius, L.; Helmersson, S. [Westinghouse Electric Sweden AB, SE-72163 Vaesteraas (Sweden); Massih, A.R. [Quantum Technologies AB, SE-75183 Uppsala and Malmoe Univ., Malmoe (Sweden); Schrire, D. [Vattenfall Braensle, SE-162 87 Stockholm (Sweden); Kaellstroem, R. [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden); Wikmark, G. [Westinghouse Electric Co., Columbia, SC 29223 (United States); Hellwig, C. [Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland)

    2007-07-01

    Some basic attributes of light water reactor fuel performance, determined by measurements, are evaluated. In particular, data on fuel volume swelling, cladding creep/growth, fission product gas release and cladding deformation due to pellet-clad mechanical interaction of rods irradiated in power reactors to rod burnups up to about 70 MWd/kgU are presented and appraised. A thermal fuel matrix swelling caused by fission products shows a linear increase in the fuel volume fraction with burnup up to 70 MWd/kgU with a mean rate of 0.76% per 10 MWd/kgU at a best-estimate level. Cladding hoop strain data due to in-reactor creep as a function of burnup from 15 to 70 MWd/kgU for pressurized water reactor (PWR) rods and from 5 to 50 MWd/kgU for boiling water reactor (BWR) rods are presented. The maximum measured cladding creep-down hoop strain in the considered BWR rods is {epsilon}{sub {theta}} {approx_equal} -0.5% and in the PWR rods {epsilon}{sub {theta}} {approx_equal} -1.25%. Rod growth data on BWR and PWR rods as a function of burnup are presented and discussed. Rod internal free volume data, measured and calculated as a function of burnup, are presented. Recent high burnup (52-70 MWd/kgU) fission product gas release data obtained by destructive methods are evaluated with the STAV7 computer code. Finally, slow power ramp experiments conducted at the Studsvik R2 reactor are simulated with the STAV7 code and it is observed that by accounting the contribution of fuel thermal gaseous swelling, the code describes the clad diameter increase due to pellet-clad mechanical interaction under the power bump satisfactorily. (authors)

  11. NEANDC specialists meeting on yields and decay data of fission product nuclides

    SciTech Connect

    Chrien, R.E.; Burrows, T.W. (eds.)

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

  12. Diffusivities of fission gas species in UO 2 and (U,Gd)O 2 nuclear fuels during irradiation

    Microsoft Academic Search

    M. Hirai; J. H. Davies; R. Williamson

    1995-01-01

    An experiment has been performed to determine fission gas diffusion coefficients in gadolinia-doped UO2 nuclear fuels under irradiation. Small specimens of UO2 containing 0, 4 and 8 wt% Gd2O3 were irradiated in the gas sweep rig of the DIDO reactor and fission gas release rates were measured as a function of temperature and sweep gas oxygen potential.The fission gas diffusion

  13. Changes of the surface-to-volume ratio and diffusion coefficient of fission gas in fuel pellets during irradiation

    Microsoft Academic Search

    Masaki Amaya; Viktors Grismanovs; Terje Tverberg

    2010-01-01

    Short-lived fission gas release from fuel pellets during irradiation was investigated based on the experimental results of the gas-flow rigs irradiated in the Halden Heavy Water Reactor (HBWR).The release-to-birth (R\\/B) rates of short-lived fission gas were measured by means of gas-flow measurement during the irradiation experiments. Surface-to-volume (S\\/V) ratios of fuel pellets and diffusion coefficients of short-lived fission gas release

  14. From EXILL (EXogam at the ILL) to FIPPS (FIssion Product Prompt ?-ray Spectrometer)

    NASA Astrophysics Data System (ADS)

    Blanc, A.; Chebboubi, A.; de France, G.; Drouet, F.; Faust, H.; Jentschel, M.; Kessedjian, G.; Kster, U.; Leoni, S.; Materna, T.; Mutti, P.; Panebianco, S.; Sage, C.; Simpson, G.; Soldner, T.; Ur, C. A.; Urban, W.; Vancraeyenest, A.

    2015-05-01

    Within the EXILL campaign a large and efficient cluster of Ge-detectors was installed around a very well collimated neutron beam. This has allowed to carry out rather complete spectroscopic studies close to the line of stability using the (n,?) reaction. Neutron rich isotopes were produced by neutron induced fission and prompt spectroscopy was carried out. The isotope selection in this setup was based on a partially known level scheme and the use of triple coincidences. The latter is limiting the statistical sensitivity in the case of weak production yields. Based on the experiences of these campaigns we are currently developing a new setup: FIPPS (FIssion Product Prompt Spectroscopy). This setup combines a collimated neutron beam, a highly efficient cluster of Ge detectors, a gas filled magnet and auxiliary detectors. The presence of the gas filled magnet will allow us to identify fission products directly and should give access to a new quality of studies if compared to the EXILL campaign. The EXILL campaign and the FIPPS project are presented.

  15. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    DOEpatents

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  16. Fission product chemistry experiments at Sandia. [PWR; BWR

    SciTech Connect

    Elrick, R.M.; Sallach, R.A.; Brockmann, J.E.

    1982-01-01

    The High Temperature Fission Product Program, an experimental and analytical effort to produce thermodynamic and reaction-rate data on fission product species during severe LWR accidents, is reviewed. The paper discusses: (1) a facility where non-radioactive isotopes of fission products react with steam, reactor materials, and hydrogen for periods of time from seconds to hours at temperatures up to 1100/sup 0/C; (2) the calibration of a laser Raman spectrometer for identifying chemical species at temperature; and (3) recent results of vapor phase studies with CsOH as well as CsI in a 304ss or Inconel 600 system with steam and hydrogen and, in the absence of steam and hydrogen, reactions of tellurium vapor with nickel, 304ss, Inconel 600, preoxidized 304ss, preoxidized Inconel 600 and silver, reaction of iodine vapor with silver, and reaction of CsOH with iodine and with HI and of CsI with Ag and with O/sub 2/.

  17. Fission gas release and grain growth in THO-UO fuel irradiated at high temperature

    Microsoft Academic Search

    I. Goldberg; L. A. Waldman; J. F. Giovengo; W. R. Campbell

    1979-01-01

    Data are presented on fission gas release and grain growth in ThO-UO fuels irradiated as part of the LWBR fuel element development program. These data were obtained from three fuel rods that experienced peak linear power outputs ranging from 15 to 22 KW\\/ft and supplement fission gas release data previously reported for 51 rods containing ThO and ThO-UO fuel irradiated

  18. Techniques for In-Pile Measurements of Fission-Gas Release

    Microsoft Academic Search

    Akira DOI; Kiyoshi INOUE; Hiroshi HAYASHI; Yasuo OHSAWA

    1970-01-01

    The continuous release of fission-gas from ceramic fuel was measured during irradiation in the Hitachi Training Reactor. Two different types of in-core assemblies were devised. In these assemblies, U02 specimens were heated up to 2,000C by a heater rod, or else cylinder of tungsten. These assemblies can usefully serve in the study of fission-gas release rate at temperatures conducive to

  19. Reactor power history from fission product signatures

    E-print Network

    Sweeney, David J.

    2009-05-15

    safeguards are needed to detect and deter any attempts to circumvent the safeguards system. Diversion of spent nuclear fuel in order to obtain plutonium is one method for weapons production.2 Assuming that it is difficult to defeat IAEA fuel assembly... of plutonium in the fuel, a higher concentration of Pu-239 in the plutonium, or a lower concentration of Pu-240 in the plutonium.3 These and other factors are dependent on the power history of the fuel assembly and allow for greater optimization...

  20. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  1. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    SciTech Connect

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  2. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    SciTech Connect

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  3. Venting of fission products and shielding in thermionic nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Salmi, E. W.

    1972-01-01

    Most thermionic reactors are designed to allow the fission gases to escape out of the emitter. A scheme to allow the fission gases to escape is proposed. Because of the low activity of the fission products, this method should pose no radiation hazards.

  4. Engineering Report on the Fission Gas Getter Concept

    SciTech Connect

    Ecker, Lynne; Ghose, Sanjit; Gill, Simerjeet; Thallapally, Praveen K.; Strachan, Denis M.

    2012-11-01

    In 2010, the Department of Energy (DOE) requested that a Brookhaven National Laboratory (BNL)-led team research the possibility of using a getter material to reduce the pressure in the plenum region of a light water reactor fuel rod. During the first two years of the project, several candidate materials were identified and tested using a variety of experimental techniques, most with xenon as a simulant for fission products. Earlier promising results for candidate getter materials were found to be incorrect, caused by poor experimental techniques. In May 2012, it had become clear that none of the initial materials had demonstrated the ability to adsorb xenon in the quantities and under the conditions needed. Moreover, the proposed corrective action plan could not meet the schedule needed by the project manager. BNL initiated an internal project review which examined three questions: 1. Which materials, based on accepted materials models, might be capable of absorbing xenon? 2. Which experimental techniques are capable of not only detecting if xenon has been absorbed but also determine by what mechanism and the resulting molecular structure? 3. Are the results from the previous techniques useable now and in the future? As part of the second question, the project review team evaluated the previous experimental technique to determine why incorrect results were reported in early 2012. This engineering report is a summary of the current status of the project review, description of newly recommended experiments and results from feasibility studies at the National Synchrotron Light Source (NSLS).

  5. ORNL studies of fission product release under LWR severe accident conditions

    Microsoft Academic Search

    M. F. Osborne; R. A. Lorenz

    2009-01-01

    The large inventories of radioactive fission products in irradiated fuel represent the principal personnel hazards from nuclear reactors. A large fraction of the existing fission-product release data has been collected from experiments at Oak Ridge National Laboratory (ORNL). Tests of high-burnup light-water-reactor fuel, and also simulated fuel with fission-product tracers, have been conducted in an induction furnace at temperatures up

  6. Fission gases and helium gas behavior in irradiated mixed oxide fuel pin

    NASA Astrophysics Data System (ADS)

    Sato, I.; Katsuyama, K.; Arai, Y.

    2011-09-01

    Behavior of helium and fission gases in irradiated mixed oxide (MOX) fuels was investigated by pin puncture and heating tests by quantitatively measuring amounts of helium and fission gases in a fuel pin irradiated in JOYO to 50 MW d kg -1 as a whole pin average burnup. While the fission gas releases were 47% and 48% for Kr and Xe respectively, all helium generated during irradiation was released (100%), and the helium released during the heating test was derived from ?-decay after irradiation. The release profile during the heating test indicated that helium gas release onset temperature was below 1173 K at an isothermal condition, but during irradiation, the helium release behavior could be understood by taking its high diffusion coefficient into consideration. The different release behavior of helium and fission is mainly explained by their different mobility in the fuel.

  7. Fission gas release from UO{sub 2+x} in defective light water reactor fuel rods

    SciTech Connect

    Skim, Y. S.

    1999-11-12

    A simplified semi-empirical model predicting fission gas release form UO{sub 2+x} fuel to the fuel rod plenum as a function of stoichiometry excess (x) is developed to apply to the fuel of a defective LWR fuel rod in operation. The effect of fuel oxidation in enhancing gas diffusion is included as a parabolic dependence of the stoichiometry excess. The increase of fission gas release in a defective BWR fuel rod is at the most 3 times higher than in an intact fuel rod because of small extent of UO{sub 2} oxidation. The major enhancement contributor in fission gas release of UO{sub 2+x} fuel is the increased diffusivity due to stoichiometry excess rather than the higher temperature caused by degraded fuel thermal conductivity.

  8. On the role of grain boundary diffusion in fission gas release

    Microsoft Academic Search

    D. R. Olander; P. Van Uffelen

    2001-01-01

    It is generally believed that thermal fission gas release from LWR fuel occurs mainly via interconnected grain boundary bubbles. Grain boundary diffusion is not considered to be a significant mechanism. We investigated this supposition by two methods; first, by assessing the distance a gas atom can migrate in a grain boundary containing perfectly absorbing traps. For areal number densities and

  9. Release Behavior of Fission Gas from Coated Fuel Particles under Irradiation

    Microsoft Academic Search

    Kousaku FUKUDA; Touru OGAWA; Satoru KASHIMURA; Katsuichi IKAWA; Kazumi IWAMOTO; Katsumune YAMAMOTO; Tadaharu ITOH; Hideo MATSUSHIMA

    1982-01-01

    TRISO coated fuel particles for HTGR were irradiated by two sweep gas capsules in order to study the release behavior of the fission gas and try to predict the failure fraction of the particles on the basis of the measurement. For verification of the predicted failure fraction, post irradiation examination was conducted, and failure fraction in a visual inspection and

  10. JRC's on-line fission gas release monitoring system in the high flux reactor Petten

    Microsoft Academic Search

    M. Laurie; M. A. Ftterer; K. H. Appelman; J. M. Lapetite; A. Marmier; G. Berg

    For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed, dubbed the Sweep Loop Facility (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experiments. The SLF

  11. Fission gas and iodine release measured in IFA430 up to 15 GWd\\/t UO burnup. [PWR; BWR

    Microsoft Academic Search

    A. D. Appelhans; J. A. Turnbull; R. J. White

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper

  12. Product yields for the photo-fission of 209Bi with 2.5 GeV bremsstrahlung

    Microsoft Academic Search

    Haladhara Naik; Sarbjit Singh; Annareddy Venkat Raman Reddy; Vijay Kumar Manchanda; Guinyun Kim; Kyung Sook Kim; Man-Woo Lee; Srinivasan Ganesan; Devesh Raj; Hee-Seock Lee; Young Do Oh; Moo-Hyun Cho; In Soo Ko; Won Namkung

    2009-01-01

    The mass-yield distribution of fission products in the 2.5GeV bremsstrahlung-induced fission of 209Bi have been determined by using the recoil catcher and the off-line ?-spectrometry technique in the high energy electron linac at the Pohang Accelerator Laboratory. The mass-yield determination involves the measurements of cumulative yields for 32 fission products and independent yields of 17 fission products in the photo-fission

  13. On-site gamma-ray spectroscopic measurements of fission gas release in irradiated nuclear fuel.

    PubMed

    Matsson, I; Grapengiesser, B; Andersson, B

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514 keV gamma-radiation from the gaseous fission product (85)Kr captured in the fuel rod plenum volume. A submergible measuring device (LOKET) consisting of an HPGe-detector and a collimator system was utilised allowing for single rod measurements on virtually all types of BWR fuel. A FGR database covering a wide range of burn-ups (up to average rod burn-up well above 60 MWd/kgU), irradiation history, fuel rod position in cross section and fuel designs has been compiled and used for computer code benchmarking, fuel performance analysis and feedback to reactor operators. Measurements clearly indicate the low FGR in more modern fuel designs in comparison to older fuel types. PMID:16949295

  14. Biological removal of cationic fission products from nuclear wastewater.

    PubMed

    Ngwenya, N; Chirwa, E M N

    2011-01-01

    Nuclear energy is becoming a preferred energy source amidst rising concerns over the impacts of fossil fuel based energy on global warming and climate change. However, the radioactive waste generated during nuclear power generation contains harmful long-lived fission products such as strontium (Sr). In this study, cationic strontium uptake from solution by microbial cultures obtained from mine wastewater is evaluated. A high strontium removal capacity (q(max)) with maximum loading of 444 mg/g biomass was achieved by a mixed sulphate reducing bacteria (SRB) culture. Sr removal in SRB was facilitated by cell surface based electrostatic interactions with the formation of weak ionic bonds, as 68% of the adsorbed Sr(2+) was easily desorbed from the biomass in an ion exchange reaction with MgCl?. To a lesser extent, precipitation reactions were also found to account for the removal of Sr from aqueous solution as about 3% of the sorbed Sr was precipitated due to the presence of chemical ligands while the remainder occurred as an immobile fraction. Further analysis of the Sr-loaded SRB biomass by scanning electron microscopy (SEM) coupled to energy dispersive X-ray (EDX) confirmed extracellular Sr(2+) precipitation as a result of chemical interaction. In summary, the obtained results demonstrate the prospects of using biological technologies for the remediation of industrial wastewaters contaminated by fission products. PMID:21245563

  15. Fission product release from trace irradiated UO 2+ x

    NASA Astrophysics Data System (ADS)

    Mansouri, M. A.; Olander, D. R.

    1998-03-01

    Fission-product release experiments were performed on trace-irradiated specimens of LWR fuel. The behavior of Te, I, Ru, Mo and Xe were followed in stoichiometric and pre-oxidized fuel. Maintenance of the desired O/U ratio was accomplished using an in situ microbalance during annealing. Fractional releases were determined from the intensities of characteristic gamma rays of the five fission product elements. Release fractions increased linearly with the square root of anneal time and decreased slightly as the specimen thickness was increased. No theoretical model could account for the effects of these two observations simultaneously. Comparison of the appropriately time- and size-scaled release data showed modest effects of oxidation on release kinetics, with Mo and Ru being most affected. Arrhenius plots exhibited breaks in linearity in the oxidized specimens, and the activation energies of the high-temperature portions were larger than commonly observed. Release during experiments in which oxidation (by steam) took place during annealing showed practically no dependence on temperature and were systematically higher than the release fractions measured during comparable constant-stoichiometry tests.

  16. ESOL facility for the generation and radiochemical separation of short half-life fission products

    SciTech Connect

    Gehrke, R.J.; Meikrantz, D.H.; Baker, J.D.; Anderl, R.A.; Novick, V.J.; Greenwood, R.C.

    1988-01-01

    A facility has been developed at the Idaho National Engineering Laboratory (INEL) for the generation and rapid radiochemical separation of short half-life mixed fission products. This facility, referred to as the Idaho Elemental Separation On Line (ESOL), consists of electro-plated sources of spontaneously fissioning /sup 252/Cf with a helium jet transport arrangement to continuously deliver short half-life, mixed fission products to the radiochemistry laboratory for rapid, computer controlled, radiochemical separations. 18 refs., 13 figs.

  17. Target and method for the production of fission product molybdenum-99

    DOEpatents

    Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

    1987-10-26

    A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

  18. System for measuring fission-gas inventories of HTGR fuel particles

    Microsoft Academic Search

    J. M. Robbins; M. J. Kania; E. L. Jr. Long; W. T. Jr. Rainey

    1979-01-01

    A Postirradiation Gas Analyzer (PGA) has been designed and built at Oak Ridge National Laboratory for quantitatively measuring the fission- and reactive-gas inventories of irradiated High-Temperature Gas-Cooled Reactor (HTGR) fuel particles. Individual particles can be broken at ambient temperature or after heating to 2000°C, and the various gases released are quantitatively analyzed. The measure of coating integrity is determined by

  19. Fission-gas release from irradiated PWR fuel during simulated PCM-type accidents: progress report

    Microsoft Academic Search

    S. M. Gehl; M. G. Seitz; J. Rest

    1977-01-01

    Radial temperature profiles and centerline heating rates of hypothetical power-cooling-mismatch accidents were simulated in irradiated commercial pressurized-water-reactor fuel with a direct-electrical-heating technique. The fission-gas release, transient temperature histories, and transient-induced microstructural changes were determined for these experiments. Empirical relationships were developed between gas release and the maximum centerline temperature and between gas release and the maximum volume-averaged temperature gradient. Preliminary

  20. Sorption of Np and U Fission Products by Zeolite Y, Mexican Natural Erionite, and Bentonite

    Microsoft Academic Search

    M. T. Olgun; M. Solache; J. L. Iturbe; P. Bosch; S. Bulbulian

    1996-01-01

    Zeolite Y, erionite, and bentonite have been used in this work to remove Np and U fission products from aqueous solutions at various pH values. It was found that the sorption of fission products by aluminosilicates takes place by different mechanisms, mainly ion exchange, precipitation, and electrostatic surface interaction. The radio nuclides content was determined by ?-spectrometry, and x-ray diffraction

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

    Microsoft Academic Search

    1958-01-01

    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.

  2. Decontamination of actinides and fission products from stainless steel surfaces

    SciTech Connect

    Mertz, C.; Chamberlain, D.B.; Chen, L.; Conner, C.; Vandegrift, G.F. [Argonne National Lab., IL (United States); Drockelman, D.; Kaminski, M.; Landsberger, S.; Stubbins, J. [Illinois Univ., Urbana, IL (United States). Dept. of Nuclear Engineering

    1996-04-01

    Seven in situ decontamination processes were evaluated as possible candidates to reduce radioactivity levels in nuclear facilities throughout the DOE complex. These processes were tested using stainless steel coupons (Type 304) contaminated with actinides (Pu and Am) or fission products (a mixture of Cs, Sr, and Gd). The seven processes were decontamination with nitric acid, nitric acid plus hydrofluoric acid, fluoboric acid, silver(II) persulfate, hydrogen peroxide plus oxalic acid plus hydrofluoric acid, alkaline persulfate followed by citric acid plus oxalic acid, and electropolishing using nitric acid electrolyte. Of the seven processes, the nitric acid plus hydrofluoric acid and fluoboric acid solutions gave the best results; the decontamination factors for 3- to 6-h contacts at 80{degree}C were as high as 600 for plutonium, 5500 for americium, 700 for cesium, 15000 for strontium, and 1100 for gadolinium.

  3. Fission gas release and swelling model of metallic fast reactor fuel

    NASA Astrophysics Data System (ADS)

    Lee, Chan Bock; Kim, Dae Ho; Jung, Youn Ho

    2001-01-01

    A mechanistic model of fission gas release and swelling for the U-Pu-10Zr metallic fuel in a fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix), was developed. Fission gas bubbles are assumed to nucleate isotropically from gas atoms in the metallic fuel matrix since they could nucleate at both the grain boundaries and the phase boundaries, which are randomly distributed inside the grain. Bubbles can grow to larger sizes by gas atom diffusion and coalescence with other bubbles. When bubble density or swelling due to bubbles reaches a threshold value, bubbles are assumed to be interconnected to each other to form an open channel to the external free space and then the fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are successively released through the open bubbles. The GRSIS model was validated by comparison with the irradiation test results of U-Pu-10Zr fuels as well as parametric studies of the key variables in the model.

  4. Early fission-gas behavior in oxide fuel: escape vs trapping

    SciTech Connect

    Cordoliani, V.; Olander, D. [Dept. of Nuclear Engineering, University of California, Berkeley CA 94707 (United States)

    2007-07-01

    A Monte Carlo code has been developed to investigate the behavior of fission-gas on the grain boundaries of UO{sub 2} fuel in the early stages of irradiation. Gas atoms arriving at the grain boundary from the grain interior undergo a random walk until they meet one of four fates: 1. nucleate a bubble by jumping next to another atom; 2. trapped by an existing bubble; 3. escape at the termination of the grain boundary at a free surface; 4. re-solved by collision with a fission fragment intersecting the grain boundary. At issue is the relative importance of escape (release) and trapping by intergranular bubbles. The switch of the dominant mechanism occurs at a critical burnup when the rate of trapping equals the arrival rate of gas atoms on the grain boundary. The critical burnup depends on fission rate and temperature, but is always less than 0.003 MWd/kgU. A very low rate of fission-gas release persists beyond the critical burnup; gas atoms from the grain interior that arrive at the grain boundary immediately adjacent to the escape surface have a small but finite probability of avoiding trapping and instead are released. The code calculates the intergranular bubble-size distribution with attention paid to the effect of re-solution. (authors)

  5. HEALTH HAZARDS FROM FISSION PRODUCTS AND FALLOUT. II. GAMMA RADIATION FROM NUCLEAR WEAPONS FALLOUT

    Microsoft Academic Search

    Bjornerstedt

    1960-01-01

    Methods of estimating the gamma radiation from fallout and fission ; products are discussed. (Gamma spectra and 0.1 Mev energy intervals for 101 ; fissions were calculated for the following times after fission: 1. 2. 5. and 10 ; hours; years. Fifty energy groups of width 1.300 to 1.399 Mev were used, ; covering the range 0.0 to 5.0 Mev.

  6. Isotopic product distributions in the near symmetric mass region in proton induced fission of 238U

    Microsoft Academic Search

    P. P. Jauho; A. Jokinen; M. Leino; J. M. Parmonen; H. Penttilae; J. yst; K. Eskola; V. A. Rubchenya

    1994-01-01

    We have studied fission product yields using 19.8 MeV proton induced fission of a thin 238U target and the on-line mass separator IGISOL. The nonselectivity of the separation method used with respect to Z has allowed accurate determination of the yields of symmetric fission for the first time. The cumulative yields for the elements from Z=40 (Zr) up to Z=47

  7. Mass-yield distribution of fission products from photo-fission of nat Pb induced by 2.5 GeV bremsstrahlung

    Microsoft Academic Search

    Haladhara Naik; Sarbjit Singh; Ashok Goswami; Vijay Kumar Manchanda; S. V. Suryanarayana; Devesh Raj; Srinivasan Ganesan; Kyung Sook Kim; Man Woo Lee; Guinyun Kim; Moo-Hyun Cho; In Soo Ko; Won Namkung

    2011-01-01

    .The mass-yield distribution of fission products in the 2.5GeV bremsstrahlung-induced fission of natPb has been measured by using a recoil catcher and an off-line -ray spectrometric technique in the 2.5GeV electron linac at the Pohang Accelerator Laboratory. The determination of mass-yield\\u000a distribution involves measurement of cumulative yields of 27 fission products and independent yields of 15 fission products\\u000a in the

  8. Yields of Short-lived Fission-Products of ^235U

    NASA Astrophysics Data System (ADS)

    Tipnis, S. V.; Campbell, J. M.; Couchell, G. P.; Li, S.; Nguyen, H. V.; Pullen, D. J.; Seabury, E. H.; Schier, W. A.

    1996-10-01

    Delayed gamma spectra from ^235U(n_th, f) fission-products were measured over delay times ranging from 0.4 to 7500 s using an HPGe detector enclosed in a NaI(Tl) Compton suppression annulus. Use of beta-gamma coincidence for background reduction, and a helium-jet/moving tape arrangement to rapidly transport the fission products to the detector allowed for a precise measurement of the delay time after fission. Yields of the individual gamma lines were compared with the intensity values listed in the Nuclear Data Sheets. From the relative intensities of the gamma lines, the relative independent and cumulative yields of the precursor isotopes were calculated using the Bateman equations and compared with the values listed in the ENDF/B-VI fission-product data base. Charge-mass complementarity was used to estimate the elemental yields of the unmeasured fission-products.

  9. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    NASA Astrophysics Data System (ADS)

    Barber, Duncan Henry

    During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A benchmark calculation demonstrates the improvement in agreement of the total inventory of those chemical elements included in the RMC fuel model to an ORIGEN-S calculation. ORIGEN-S is the Oak Ridge isotope generation and depletion computer program. The Gibbs energy minimizer requires a chemical database containing coefficients from which the Gibbs energy of pure compounds, gas and liquid mixtures, and solid solutions can be calculated. The RMC model of irradiated uranium dioxide fuel has been converted into the required format. The Gibbs energy minimizer has been incorporated into a new model of fission-product vaporization from the fuel surface. Calculated release fractions using the new code have been compared to results calculated with SOURCE IST 2.0P11 and to results of tests used in the validation of SOURCE 2.0. The new code shows improvements in agreement with experimental releases for a number of nuclides. Of particular significance is the better agreement between experimental and calculated release fractions for 140La. The improved agreement reflects the inclusion in the RMC model of the solubility of lanthanum (III) oxide (La2O3) in the fuel matrix. Calculated lanthanide release fractions from earlier computer programs were a challenge to environmental qualification analysis of equipment for some accident scenarios. The new prototype computer program would alleviate this concern. Keywords: Nuclear Engineering; Material Science; Thermodynamics; Radioactive Material, Gibbs Energy Minimization, Actinide Generation and Depletion, FissionProduct Generation and Depletion.

  10. Analysis and numerical optimization of gas turbine space power systems with nuclear fission reactor heat sources

    Microsoft Academic Search

    Albert J. Juhasz

    2005-01-01

    A new three objective optimization technique is developed and applied to find the operating conditions for fission reactor heated Closed Cycle Gas Turbine (CCGT) space power systems at which maximum efficiency, minimum radiator area, and minimum total system mass is achieved. Such CCGT space power systems incorporate a nuclear reactor heat source with its radiation shield; the rotating turbo-alternator, consisting

  11. Analysis of fission gas release kinetics by on-line mass spectrometry

    Microsoft Academic Search

    Y. Zerega; C. Reynard-Carette; D. Parrat; M. Carette; B. Brkic; A. Lyoussi; G. Bignan; A Janulyte; J. Andre; Y. Pontillon; G. Ducros; S. Taylor

    2011-01-01

    The release of fission gas (Xe and Kr) and helium out of nuclear fuel materials in normal operation of a nuclear power reactor can constitute a strong limitation of the fuel lifetime. Moreover, radioactive isotopes of Xe and Kr contribute significantly to the global radiological source term released in the primary coolant circuit in case of accidental situations accompanied by

  12. Instabilities in fissioning plasmas as applied to the gas-core nuclear rocket-engine

    NASA Technical Reports Server (NTRS)

    1973-01-01

    The compressional wave spectrum excited in a fissioning uranium plasma confined in a cavity such as a gas cored nuclear reactor, is studied. Computer results are presented that solve the fluid equations for this problem including the effects of spatial gradients, nonlinearities, and neutron density gradients in the reactor. Typically the asymptotic fluctuation level for the plasma pressure is of order 1 percent.

  13. Studies of Ceramic Fuels with Use of Fission Gas Release Loop, (IV)

    Microsoft Academic Search

    Muneo HANDA; Shigeru YAMAGISHI; Takeshi FUKUDA; Koreyuki SHIBA; Yoshihisa TAKAHASHI; Takaaki TANIFUJI; Shunz? ?MORI; Akinori KOND?

    1974-01-01

    In-pile release of fission gas from sintered UC pellets in the presence of 8230 ppm of water vapor in the He sweep gas was measured over the temperature range of 1601,000C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established

  14. Fission product release and fuel behavior of irradiated light water reactor fuel under severe accident conditions

    SciTech Connect

    Allen, M.D.; Stockman, H.W.; Reil, K.O. (Sandia National Labs., Albuquerque, NM (United States)); Fisk, J.W. (Tills (Jack) and Associates, Inc., Albuquerque, NM (United States))

    1991-11-01

    The annular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain time-resolved data on the release of fission products from irradiated fuels under well-controlled light water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST-1 was conducted in a highly reducing environment at a system pressure of approximately 0.19 MPa, and at maximum fuel temperatures of about 2490 K. The data will be used for the development and validation of mechanistic fission product release computer codes such as VICTORIA.

  15. Identification of Radial Position of Fission Gas Release in High-Burnup Fuel Pellets under RIA Conditions

    Microsoft Academic Search

    Hideo SASAJIMA; Tomoyuki SUGIYAMA; Toshinori CHUTO; Fumihisa NAGASE; Takehiko NAKAMURA; Toyoshi FUKETA

    2010-01-01

    The radial positions of fission gas release (FGR) in high-burnup fuel pellets were examined after pulseirradiations that simulated reactivity-initiated accident (RIA) conditions in the Nuclear Safety Research Reactor (NSRR). The molar ratio of xenon (Xe) to krypton (Kr) (Xe\\/Kr ratio) in the released gas showed that fission gas was released from the entire region of the pellets of the examined

  16. Evolution of fission-gas-bubble-size distribution in recrystallized U-10Mo nuclear fuel

    NASA Astrophysics Data System (ADS)

    Rest, J.

    2010-12-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles, used to characterize fission-gas bubble development in U-Mo alloy fuel with burnup limited to less than 10 at.% U in order to capture the fuel swelling stage prior to irradiation-induced recrystallization, is extended to recrystallized fuel at a burnup of 16 at.% U. During recrystallization the grain size is transformed from micron to sub-micron sizes. The intergranular bubble-size distribution post-recrystallization is found to evolve with similar kinetics and morphology to that pre-recrystallization with any differences primarily due to gas content and initial and/or boundary conditions (e.g., fuel microstructure). The predictions of the theory are compared with measured bubble-size distributions in pre and post recrystallized U-10Mo alloy fuel.

  17. Zirconium and fission product management in the ALSEP process

    SciTech Connect

    Lumetta, G.J.; Carter, J.C.; Niver, C.M. [Pacific Northwest National Laboratory: P.O. Box 999, MSIN P7-25, Richland, WA 99352 (United States)

    2013-07-01

    Solvent extraction systems that combine neutral donor extractants and acidic extractants are being investigated to provide a single process solvent for separating Am and Cm from acidic high-level liquid waste, including their separation from the trivalent lanthanides. This approach of combining extractants is collectively referred to as the Actinide-Lanthanide Separation (ALSEP) process. Managing Zr and other fission products is one of the critical factors in developing the ALSEP process. In this work, a strategy has been developed in which Zr(IV) is extracted into the process solvent, then it is stripped from the solvent after the actinides have been selectively stripped. The ALSEP solvent contains a bifunctional neutral donor extractant that extracts the minor actinides and the trivalent lanthanides (Ln) from nitric acid media. In this work, two such extractants were considered: N,N,N',N'- tetraoctyl-diglycolamide (TODGA) and N,N,N',N'-tetra(2- ethylhexyl)diglycolamide (T2EHDGA). Molybdenum is strongly extracted into ALSEP solvents. Scrubbing the solvent with a citrate buffer before the actinide stripping step effectively removes Mo. Distribution ratios for Ru and Fe are low for extraction from HNO{sub 3}, so these components can easily be routed to the high-level waste raffinate. (authors)

  18. Data summary report for fission product release Test VI-7

    SciTech Connect

    Osborne, M.F.; Lorentz, R.A.; Travis, J.R.; Collins, J.L.; Webster, C.S. [Oak Ridge National Lab., TN (United States)

    1995-05-01

    Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% for Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.

  19. Fission products from the damaged Fukushima reactor observed in Hungary.

    PubMed

    Bihari, rpd; Dezs?, Zoltn; Bujts, Tibor; Manga, Lszl; Lencss, Andrs; Dombvri, Pter; Csige, Istvn; Ranga, Tibor; Mogyorsi, Magdolna; Veres, Mihly

    2014-01-01

    Fission products, especially (131)I, (134)Cs and (137)Cs, from the damaged Fukushima Dai-ichi nuclear power plant (NPP) were detected in many places worldwide shortly after the accident caused by natural disaster. To observe the spatial and temporal variation of these isotopes in Hungary, aerosol samples were collected at five locations from late March to early May 2011: Institute of Nuclear Research, Hungarian Academy of Sciences (ATOMKI, Debrecen, East Hungary), Paks NPP (Paks, South-Central Hungary) as well as at the vicinity of Aggtelek (Northeast Hungary), Tapolca (West Hungary) and Btaapti (Southwest Hungary) settlements. In addition to the aerosol samples, dry/wet fallout samples were collected at ATOMKI, and airborne elemental iodine and organic iodide samples were collected at Paks NPP. The peak in the activity concentration of airborne (131)I was observed around 30 March (1-3 mBq m(-3) both in aerosol samples and gaseous iodine traps) with a slow decline afterwards. Aerosol samples of several hundred cubic metres of air showed (134)Cs and (137)Cs in detectable amounts along with (131)I. The decay-corrected inventory of (131)I fallout at ATOMKI was 2.10.1 Bq m(-2) at maximum in the observation period. Dose-rate contribution calculations show that the radiological impact of this event at Hungarian locations was of no considerable concern. PMID:24437973

  20. Baseline Glass Development for Combined Fission Products Waste Streams

    SciTech Connect

    Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

    2009-06-29

    Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.[1] Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.[2-5] Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.

  1. Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions

    SciTech Connect

    Pigni, Marco T [ORNL; Francis, Matthew W [ORNL; Gauld, Ian C [ORNL

    2015-01-01

    A recent implementation of ENDF/B-VII. independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear scheme in the decay sub-library that is not reflected in legacy fission product yield data. Recent changes in the decay data sub-library, particularly the delayed neutron branching fractions, result in calculated fission product concentrations that are incompatible with the cumulative fission yields in the library, and also with experimental measurements. A comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron induced fission for 235,238U and 239,241Pu in order to provide a preliminary assessment of the updated fission product yield data consistency. These updated independent fission product yields were utilized in the ORIGEN code to evaluate the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. An important outcome of this work is the development of fission product yield covariance data necessary for fission product uncertainty quantification. The evaluation methodology combines a sequential Bayesian method to guarantee consistency between independent and cumulative yields along with the physical constraints on the independent yields. This work was motivated to improve the performance of the ENDF/B-VII.1 library in the case of stable and long-lived cumulative yields due to the inconsistency of ENDF/B-VII.1 fission p;roduct yield and decay data sub-libraries. The revised fission product yields and the new covariance data are proposed as a revision to the fission yield data currently in ENDF/B-VII.1.

  2. FPTRAN: A Volatile Fission Products and Structural Materials Transport Code for SCDAP/RELAP5

    SciTech Connect

    Honaiser, Eduardo [Brazilian Navy Technological Center, R. Professor Lineu Prestes, 2468, Sao Paulo, SP (Brazil); Anghaie, Samim [Innovative Space Power and Propulsion Institute, 2800 SW Archer Rd. Bldg, 554, P.O. Box 116502, University of Florida, Gainesville, FL, 32611-6502 (United States)

    2004-07-01

    The fission products behavior in reactor coolant systems (RCS) is divided in the fission products release from the fuel, transport through the piping system, and the chemistry of the several materials present in a LWR. The transport poses significant difficulty for the implementation, due to the complexity in the treatment of the system of equations generated for the solution, as well as the difficulties in the modeling of certain phenomena. This paper presents the FPTRAN code, which was incorporated to SCDAP/RELAP5, and initially tested satisfactorily. FPTRAN does the calculation of the transport of fission products in RCS, estimating the amount of material being deposited over the pipes, and the amount released to the containment, once a source of released material (fission products and structural materials) to the piping system is provided. (authors)

  3. New Fission-Product Waste Forms: Development and Characterization

    SciTech Connect

    Alexandra Navrotsky

    2010-07-30

    Research performed on the program New Fission Product Waste Forms: Development and Characterization, in the last three years has fulfilled the objectives of the proposal which were to 1) establish ceramic waste forms for disposing of Cs, Sr and minor actinides, 2) fully characterize the phase relationships, structures and thermodynamic and kinetic stabilities of promising waste forms, 3) establish a sound technical basis for understanding key waste form properties, such as melting temperatures and aqueous durability, based on an in-depth understanding of waste form structures and thermochemistry, and 4) establish synthesis, testing, scaleup and commercialization routes for wasteform implementation through out in-kind collaborations. In addition, since Cs and Sr form new elements by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products were discovered using non-radioactive analogues. Collaborations among researchers from three institutions, UC Davis, Sandia National Laboratories, and Shott Inc., were formed to perform the primary work on the program. The unique expertise of each of the members in the areas of waste form development, structure/property relationships, hydrothermal and high temperature synthesis, crystal/glass production, and thermochemistry was critical to program success. In addition, collaborations with the Brigham Young Univeristy, Ben Gurion University, and Los Alamos National Laboratory, were established for standard entropies of ceramic waste forms, sol-gel synthesis, and high temperature synthesis. This work has had a significant impact in a number of areas. First, the studies of the thermodynamic stability of the mineral analogues provided an important technical foundation for assessment the viability of multicomponent oxide phases for Cs and Sr removal. Moreover, the thermodynamic data discovered in this program established information on the reaction pathways for the potential reaction products. The phase equilibria and thermodynamics involving the intermediates in the decay process in this program will assist in selection of the best process for Cs or Sr immobilization. In addition, data from the study can be used to develop engineering solutions for potential process upsets. Second, the glass crystal stability of multicomponent oxide phases that were representative silicates on this program is highly distinguishable for mother compounds and decay products, thus providing a fundamental understanding on the separate effects from chemistry and from radiation. Finally, we have developed a foundation for understanding chemistry-structure-energetics relationships in titanosilicates that can be used to develop more effective materials.

  4. Lumped fission product neutron cross sections based on ENDF\\/BV for fast reactor analysis

    Microsoft Academic Search

    J. R. Liaw; H. Henryson

    1983-01-01

    The development and evaluation of a lumped fission product neutron cross-section library based on ENDF\\/B-V data suitable for fast reactor applications have been completed. Both one- and two-lump models have been investigated in detail. Fission product inventories at various burnup steps were calculated by the EPRI-CINDER-2 code and used as weighting functions for lumping. This paper addresses several important issues

  5. Fission product release measured during fuel damage tests at the Power Burst Facility

    Microsoft Academic Search

    D. J. Osetek; J. K. Hartwell; K. Vinjamuri; A. W. Cronenberg

    1985-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid quench and slow cooldown, low and high (36 GWd\\/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product

  6. TIME-DEPENDENT FISSION-PRODUCT POISONS IN U-235 AND NATURAL URANIUM FUELS

    Microsoft Academic Search

    M. R. Stuart; T. R. England

    1962-01-01

    Fission-product poison in the core of the PWR seedblanket reactor was ; calculated. The fission-product poison in U²³⁵ is compared with that of ; natural uranium and Pu²³⁹ The effects of a power-dependent poison are ; described. The aggregate poison stability is discussed. The values of ; calculated and experimental poisons are compared. The results are based on ; explicit

  7. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  8. High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

    Microsoft Academic Search

    F. Lemoine

    1997-01-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning

  9. Angular distribution of products of ternary nuclear fission induced by cold polarized neutrons

    SciTech Connect

    Bunakov, V. E., E-mail: bunakov@vb13190.spbu.edu; Kadmensky, S. G., E-mail: kadmensky@phys.vsu.ru; Kadmensky, S. S. [Voronezh State University (Russian Federation)

    2008-11-15

    Within quantum fission theory, angular distributions of products originating from the ternary fission of nuclei that is induced by polarized cold and thermal neutrons are investigated on the basis of a non-evaporative mechanism of third-particle emission and a consistent description of fission-channel coupling. It is shown that the inclusion of Coriolis interaction both in the region of the discrete and in the region of the continuous spectrum of states of the system undergoing fission leads to T-odd correlations in the aforementioned angular distributions. The properties of the TRI and ROT effects discovered recently, which are due to the interference between the fission amplitudes of neutron resonances, are explored. The results obtained here are compared with their counterparts from classic calculations based on the trajectory method.

  10. Fission gas release from ThO and ThO--UO fuels (LWBR development program)

    Microsoft Academic Search

    I. Goldberg; G. L. Spahr; L. S. White; L. A. Waldman; J. F. Giovengo; P. L. Pfennigwerth; J. Sherman

    1978-01-01

    Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO or ThO-UO fuel pellets, with UO compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71

  11. Fission gas release and volume diffusion enthalpy in UO 2 irradiated at low and high burnup

    Microsoft Academic Search

    J. P Hiernaut; C. Ronchi

    2001-01-01

    Samples of UO2, irradiated in LWRs at burnups from 25000 to 95000 MWd\\/t at in-pile temperatures below 800 K, were submitted to Knudsen-effusion experiments. In addition to the equilibrium vapour pressure of non-volatile species, release of fission gas was measured as a function of temperature and time up to approximately 2700 K, where complete vaporisation of the sample was eventually

  12. In-Pile Fission Gas Release from UC at Temperature around 2,000C

    Microsoft Academic Search

    Hirotaka FURUYA; Junji KOMATSU; Tadashi MUTO; Akira DOI; Kiyoshi INOUE

    1971-01-01

    In-pile fission gas release from nearly stoichiometric sintered uranium monocarbide was measured in the temperature range from 1,650 to 2,030C. The result was analyzed on the basis of the diffusion model, from which the diffusion coefficient of krypton in uranium monocarbide was derived asD=1.610 exp(56,400\\/RT) cm\\/sec.These values were compared with the result obtained with UO2 in the same irradiation facility.

  13. A Fission Gas Release Model for High-Burnup LWR ThO-UO Fuel

    Microsoft Academic Search

    Yun Long; Yi Yuan; Mujid S. Kazimi; Ronald G. Ballinger; Edward E. Pilat

    2002-01-01

    Fission gas release in thoria-urania fuel has been investigated by creating a specially modified FRAPCON-3 code. Because of the reduced buildup of ²³⁹Pu and a flatter distribution of ²³³U, the new model THUPS (Thoria-Urania Power Shape) was developed to calculate the radial power distribution, including the effects of both plutonium and ²³³U. Additionally, a new porosity model for the rim

  14. Fission gas release and swelling in uraniumplutonium mixed nitride fuels

    Microsoft Academic Search

    Kosuke Tanaka; Koji Maeda; Kozo Katsuyama; Masaki Inoue; Takashi Iwai; Yasuo Arai

    2004-01-01

    Two uraniumplutonium mixed nitride, (U,Pu)N, fuel pins with different He-gap width were irradiated at a linear heating rate 75 kW\\/m to 4.3% FIMA in the experimental fast reactor JOYO, and nondestructive and destructive post irradiation examinations were carried out. Fission gas release rates were about 3.3% and 5.2%, and swelling rates were about 1.8% and 1.6%\\/% FIMA. From the radial

  15. Gas production during refuse decomposition

    Microsoft Academic Search

    G. J. Farquhar; F. A. Rovers

    1973-01-01

    Gas production in sanitary landfills is a subject of much concern because of the potential hazards of CH4 combustion and of groundwater contamination by CO2. This study investigated the pattern of sanitary landfill gas production and the factors which affect it.

  16. Detecting special nuclear materials in containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B. (Oakland, CA); Prussin, Stanley G. (Kensington, CA)

    2007-10-02

    A method and a system for detecting the presence of special nuclear materials in a container. The system and its method include irradiating the container with an energetic beam, so as to induce a fission in the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  17. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    SciTech Connect

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  18. Analysis of fission gas release in LWR fuel using the BISON code

    SciTech Connect

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  19. Continuous fission-product monitor system at Oyster Creek. Final report

    SciTech Connect

    Collins, L.L.; Chulick, E.T.

    1980-10-01

    A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for the period from April 1979 through January 1980, the end of cycle 8 for the Oyster Creek plant. Accumulated spectra, an average of three a day, were stored on magnetic disk and subsequently analyzed for fisson products, Because of difficulties in measuring absolute detector efficiency, quantitative fission product concentrations in the coolant could not be determined. Data for iodine fission products are reported as a function of time. The data indicate the existence of fuel defects in the Oyster Creek core during cycle 8.

  20. The rate of decay of fresh fission products from a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Dolan, David J.

    Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

  1. Migration of fission products at the Nevada Test Site: Detection with an isotopic tracer

    SciTech Connect

    Thompton, J.L.; Gilmore, J.S. (Los Alamos National Lab., NM (USA))

    1989-01-01

    Researchers at Los Alamos National Laboratory are studying the migration of fission products away from explosion cavities formed by underground nuclear tests at the Nevada Test Site. In some cases, the isotopic composition of the fission products or activation products associated with a particular test are distinctive and we may identify them many years after the event. In this paper we describe a case in which we used rhodium isotopes to identify the source of radioactive material that had moved some 350 m from the explosion site. 4 refs., 2 figs., 2 tabs.

  2. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    SciTech Connect

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  3. Augmentation of ENDF/B fission product gamma-ray spectra by calculated spectra

    SciTech Connect

    Katakura, J. (Japan Atomic Energy Research Inst., Tokai-mura, Naka-gun, Ibaraki-ken (Japan)) [Japan Atomic Energy Research Inst., Tokai-mura, Naka-gun, Ibaraki-ken (Japan); England, T.R. (Los Alamos National Lab., NM (United States)) [Los Alamos National Lab., NM (United States)

    1991-11-01

    Gamma-ray spectral data of the ENDF/B-V fission product decay data file have been augmented by calculated spectra. The calculations were performed with a model using beta strength functions and cascade gamma-ray transitions. The calculated spectra were applied to individual fission product nuclides. Comparisons with several hundred measured aggregate gamma spectra after fission were performed to confirm the applicability of the calculated spectra. The augmentation was extended to a preliminary ENDF/B-VI file, and to beta spectra. Appendix C provides information on the total decay energies for individual products and some comparisons of measured and aggregate values based on the preliminary ENDF/B-VI files. 15 refs., 411 figs.

  4. Trapping and diffusion of fission products in ThO2 and CeO2

    SciTech Connect

    Xiao, Haiyan [University of Tennessee, Knoxville (UTK); Zhang, Yanwen [ORNL; Weber, William J [ORNL

    2011-01-01

    The trapping and diffusion of Br, Rb, Cs and Xe in ThO2 and CeO{sub 2} have been studied using an Ab Initio total energy method in the local-density approximation of density functional theory. Fission products incorporated in cation mono-vacancy, cation-anion di-vacancy and Schottky defect sites are found to be stable, with the cation mono-vacancy being the preferred site in most cases. In both oxides, Rb and Cs are the most likely to be trapped, and Xe is more difficult to incorporate than other fission products. The energy barriers for migration of each species in ThO{sub 2} and CeO{sub 2} are also calculated. Alkali metals are relatively more mobile than other fission products, and bromine is the least mobile.

  5. High temperature fission product chemistry and transport in steam. [PWR; BWR

    SciTech Connect

    Elrick, R.M.; Sallach, R.A.

    1982-01-01

    The High Temperature Fission Product Program, an experimental and analytical effort to produce thermodynamic and reaction-rate data on fission product species during severe LWR accidents, is reviewed. The paper discusses: (1) a facility where non-radioactive isotopes of fission products react with steam, reactor materials, and hydrogen for periods of time from seconds to hours at temperatures up to 1100/sup 0/C; and (2) recent results of vapor phase studies with CsOH as well as CsI in a 304ss or Inconel 600 system with steam and hydrogen and, in the absence of steam and hydrogen, reactions of tellurium vapor with nickel, 304ss, Inconel 600, preoxidized 304ss, preoxidized Inconel 600 and silver, iodine vapor with silver, CsOH with iodine and with HI and CsI with Ag and with O/sub 2/.

  6. Analysis of transient fission gas release and swelling in oxide fuel

    SciTech Connect

    Gruber, E.E.

    1985-01-01

    Calculations of fission-gas behavior have been carried out with an updated version of the FRAS3 code for the FGR-40 series of transient tests of irradiated PNL-10 fuel. This same series of tests was used in an earlier evaluation study with a preliminay version of the code. While that study provided positive support for the modeling approach, it also indicated deficiencies in some areas. Although a number of improvements have been implemented in the current version of the code, this study examines the effect of an explicit treatment of bubble growth within the grains, including the effect of vacancy depletion caused by the competition of overpressured bubbles for available vacancies. The result is a reduction by as much as 90% in the predicted swelling, accompanied by an increase in transfer of gas from the grains to boundaries. Both swelling and gas release predictions are brought into much better agreement with the observed values.

  7. A METHOD FOR THE STUDY AND CORRELATION OF FISSION-GAS-RELEASE BEHAVIOR OF FUEL MATERIALS DURING IRRADIATION

    Microsoft Academic Search

    G. E. Raines; C. W. Townley; S. D. Beck; W. H. Goldthwaite

    1961-01-01

    C-coated and SiC-coated 1.5-in.-diameter graphite spheres fueled with ; natural UC and UC\\/sub2\\/ particles were irradiated at 130 to 1900 deg at ; relatively low flux (<5 x 10\\/sup 12 n\\/(cm²(sec)l in an in-pile study of ; fission-gas release. A stream of He flowed continuously over the spheres and was ; sampled periodically for fission gases. Five species Kr\\/sup 85m

  8. Diffusion of fission products and radiation damage in SiC

    NASA Astrophysics Data System (ADS)

    Malherbe, Johan B.

    2013-11-01

    A major problem with most of the present nuclear reactors is their safety in terms of the release of radioactivity into the environment during accidents. In some of the future nuclear reactor designs, i.e. Generation IV reactors, the fuel is in the form of coated spherical particles, i.e. TRISO (acronym for triple coated isotropic) particles. The main function of these coating layers is to act as diffusion barriers for radioactive fission products, thereby keeping these fission products within the fuel particles, even under accident conditions. The most important coating layer is composed of polycrystalline 3C-SiC. This paper reviews the diffusion of the important fission products (silver, caesium, iodine and strontium) in SiC. Because radiation damage can induce and enhance diffusion, the paper also briefly reviews damage created by energetic neutrons and ions at elevated temperatures, i.e. the temperatures at which the modern reactors will operate, and the annealing of the damage. The interaction between SiC and some fission products (such as Pd and I) is also briefly discussed. As shown, one of the key advantages of SiC is its radiation hardness at elevated temperatures, i.e. SiC is not amorphized by neutrons or bombardment at substrate temperatures above 350 C. Based on the diffusion coefficients of the fission products considered, the review shows that at the normal operating temperatures of these new reactors (i.e. less than 950 C) the SiC coating layer is a good diffusion barrier for these fission products. However, at higher temperatures the design of the coated particles needs to be adapted, possibly by adding a thin layer of ZrC.

  9. THE EFFECT OF CLOSED POROSITY ON THE ACCURACY OF THE MODEL FOR FISSION-GAS RELEASE FROM UO

    Microsoft Academic Search

    Rosenthal

    1959-01-01

    The presence in U0 compacts of closed pores capable of retaining ; fission-product gases introduces a factor which is not considered in the normal ; treatment of fissiongas behavior. An analysis indicates that if a \\

  10. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light

  11. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    SciTech Connect

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000/sup 0/C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab.

  12. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    A preliminary analysis of the re-evaporization of volatile fission product from a boiling water reactor (BWR) cooling system following a core meltdown accident in which the core debris penetrates the reactor vessel has been performed. The BWR analyzed has a Mark I containment and the accident sequence was a station blackout transient. This work was performed as part of the phenomenological uncertainty study of the Quantification and Uncertainty Analysis of Source Terms for Severe Accidents program at Brookhaven National Laboratory. Fission product re-evaporization was identified as one of the important issues in the Reactor Risk Reference Document.

  13. Chemical forms of solid fission products in the irradiated uraniumplutonium mixed nitride fuel

    NASA Astrophysics Data System (ADS)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

  14. Phase-field simulations of intragranular fission gas bubble evolution in UO2 under post-irradiation thermal annealing

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert O.; Gao, Fei; Sun, Xin

    2013-05-15

    Fission gas bubble is one of evolving microstructures, which affect thermal mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phasefield model was developed to describe the evolution kinetics of intra-granular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nuclei size of the gas bubble and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter is predicted which is in a good agreement with experimental data.

  15. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    NASA Astrophysics Data System (ADS)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Bene, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    Samples of irradiated MOX fuel of 44.5 GWd/tHM mean burn-up were prepared by core drilling at three different radial positions of a fuel pellet. They were subsequently heated in a Knudsen effusion mass spectrometer up to complete vaporisation of the sample (2600 K) and the release of fission gas (krypton and xenon) as well as helium was measured. Scanning electron microscopy was used in parallel to investigate the evolution of the microstructure of a sample heated under the same condition up to given key temperatures as determined from the gas release profiles. A clear initial difference for fission gas release and microstructure was observed as a function of the radial position of the samples and therefore of irradiation temperature. A good correlation between the microstructure evolution and the gas release peaks could be established as a function of the temperature of irradiation and (laboratory) heating. The region closest to the cladding (0.58 < r/r0 < 0.96), designated as sample type A in Fig. 1. It represents the "cooler" part of the fuel pellet. The irradiation temperatures (Tirrad) in this range are from 854 to 1312 K (?T: 458 K). The intermediate radial zone of the pellet (0.42 < r/r0 < 0.81), designated sample type B in Fig. 1, has a Tirrad ranging from 1068 to 1434 K (?T: 365 K). The central zone of the pellet (0.003 < r/r0 < 0.41), designated sample type C in Fig. 1, which was close to the hottest part of the pellet, has a Tirrad ranging from 1442 to 1572 K (?T: 131 K). The sample irradiation temperatures were determined from the calculated temperature profile (exponential function) knowing the core temperature of the fuel (1573 K) [11], the standard temperature for this type of fuel at the inner side of the cladding (800 K). The average burnup was calculated with TRANSURANUS code [12] and the PA burnup is the average burnup multiplied by the ratio of the fissile Pu concentration in PA over average fissile Pu concentration in fuel [11]. Calculated burnups correspond reasonably well with measurement of Walker et al. [11]. All those data are shown Fig. 2.Fragments of 2-8 mg were chosen for the experiments. Since these specimens are small compared to the drilled sample size and were taken randomly, the precise radial position could not be determined, in particular the specimens of sample type, A and B could be from close radial locations.Specimens from each drilled sample type were annealed up to complete vaporisation (2600 K) at a speed of about 10 K min-1 in a Knudsen effusion mass spectrometer (KEMS) described previously [13,14]. In addition to helium and to the FGs all the species present in the vapour between 83 and 300 a.m.u. were measured during the heating. Additionally, the 85Kr isotope was analysed in a cold trap by ? and ? counting. The long-lived fission gas isotopes correspond to masses 131, 132, 134 and 136 for Xe and 83, 84, 85 and 86 for Kr. The absolute quantities of gas released from specimens of sample types A and B were also determined using the in-house built Q-GAMES (Quantitative gas measurement system), described in detail in [15].For each of the samples, fragments were also annealed and measured in the KEMS up to specific temperatures corresponding to different stages of the FGs or He release. These fragments were subsequently analysed by Scanning Electron Microscopy (SEM, Philips XL40) [16] in order to investigate the relationship between structural changes, burn-up, irradiation temperature and fission products release. SEM observations were also done on the samples before the KEMS experiments and the fracture surface appearance of the samples is shown in Fig. 3, revealing the presence of the high burnup structure (HBS) in the Pu-rich agglomerates.A summary of the 12 samples analysed by KEMS, SEM and Q-GAMES is given in Table 1. At 1300 K no clear change potentially related to gas release appears in the UM and PA. At 1450 K a beginning of grain boundaries opening can be observed as well as rounding of the grains attributed to thermal etching. A

  16. Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site

    SciTech Connect

    Bibler, N.E.; Coleman, C.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Kinard, W.F. [Charleston Coll., SC (United States). Dept. of Chemistry

    1992-10-01

    This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses. This paper presents measurements of the concentrations of 42 (98%) of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at the Savannah River Site. We analyzed the sludge with inductively coupled plasma-mass spectroscopy. The relative yields for most of the fission products agree completely with the known relative vields for the beta decay chains of the two asymmetric: branches of the slow neutron fission of U-235. The chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses explain the differences in the measured and calculated results.

  17. Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site

    SciTech Connect

    Bibler, N.E.; Coleman, C.J. (Westinghouse Savannah River Co., Aiken, SC (United States)); Kinard, W.F. (Charleston Coll., SC (United States). Dept. of Chemistry)

    1992-01-01

    This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses. This paper presents measurements of the concentrations of 42 (98%) of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at the Savannah River Site. We analyzed the sludge with inductively coupled plasma-mass spectroscopy. The relative yields for most of the fission products agree completely with the known relative vields for the beta decay chains of the two asymmetric: branches of the slow neutron fission of U-235. The chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses explain the differences in the measured and calculated results.

  18. Four-Fold Data Analysis of 252Cf Fission Products

    NASA Astrophysics Data System (ADS)

    Wang, Enhong; Brewer, N. T.; Hamilton, J. H.; Ramayya, A. V.; Hwang, J. K.; Luo, Y. X.; Rasmussen, J. O.; Zhu, S. J.; Ter-Akopian, G. M.; Oganessian, Yu. Ts.

    2014-09-01

    Prompt gamma-ray 4-fold data were built to collect 21011 ? -? -? -? quadruple- and higher-fold ? -coincidence events from the spontaneous fission of 252Cf with Gammasphere detector arrays. The nuclei 106Nb, 115Pd, 142La, 145,146Ba, 152Ce and Gd have been studied with these data. By using the new 4-fold data, we confirmed several weak tentative transitions in 106Nb, 142La, 145,146Ba, 148Ce which were observed previously from the ? -? -? triple cube. Some new transitions in 106Nb, 142La were identified by our new 4-fold data. Cascades in 145,146Ba are much clearer in four-fold data than the previous triple coincidence data. We will continue to study other nuclei by our 4-fold data with lower background than the previous triple cube.

  19. A Venturi Scrubber Installation for the Removal of Fission Products from Air

    Microsoft Academic Search

    H. S. Jordan; C. G. Welty

    1959-01-01

    ber installation for the cleaning of exhaust air contaminated with acid ; misis and mixed fission products are described in detail. It was determined that ; 20 cfm exhausted by a local slot exhaust hood would control the maximum evolution ; of gases from a 1500 ml beaker. Features of the exhaust system that were ; designed io offset the

  20. Evaluation of fission product nuclear data for 28 important nuclides. [Optical, statistical models

    Microsoft Academic Search

    S. Igarasi; S. Iijima; M. Kawai; T. Nakagawa

    1975-01-01

    Evaluation of 28 fission product nuclear data for fast reactors is performed for total, capture, elastic scattering and inelastic scattering cross sections up to 15.0 MeV. Resonance parameters as well as the data of resonance integrals are surveyed. The cross sections reproduced with these parameters are adjusted so as to fit the thermal values and are connected smoothly with the

  1. Release and transport of fission product cesium in the TMI-2 accident

    SciTech Connect

    Lorenz, R.A.; Collins, J.L.

    1986-01-01

    Approximately 50% of the fission product cesium was released from the overheated UO/sub 2/ fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that approx.62% of the released /sup 137/Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL).

  2. Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs

    SciTech Connect

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1980-11-01

    Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design.

  3. US\\/UK actinides experiment at the Dounreay PFR. 1: Fission products

    Microsoft Academic Search

    S. Raman; B. D. Murphy

    1995-01-01

    The US and the United Kingdom have been engaged in a joint research program in which samples of higher actinides were irradiated in the 600-MW Dounreay Prototype Fast Reactor in Scotland. Analytical results using mass spectrometry and radiometry for actinides and fission products are now available for the samples in Fuel Pins 1 and 2 which were irradiated for 63

  4. Evaluation of Neutron Cross Section of 27 Fission Product Nuclides Important for Fast Reactor

    Microsoft Academic Search

    Shungo IIJIMA; Tsuneo NAKAGAWA; Yasuyuki KIKUCHI; Masayoshi KAWAI; Hiroyuki MATSUNOBU; Koichi MAKI; Sin-iti IGARASI

    1977-01-01

    Results of evaluation of neutron cross sections are presented for 27 fission product nuclides selected as being most important for fast reactor calculation. The cross sections considered are total, elastic scattering, inelastic scattering and capture cross sections in the energy range from thermal to 15 MeV. Thermal and resonance cross sections were calculated from resonance parameters. The calculated thermal capture

  5. FISSION-PRODUCT SEPARATION BASED ON ROOM-TEMPERATURE IONIC LIQUIDS

    EPA Science Inventory

    The objectives of this project are (a) to synthesize new ionic liquids tailored for the extractive separation of Cs + and Sr 2+; (b) to select optimum macrocyclic extractants through studies of complexation of fission products with macrocyclic extractants and transport in new ext...

  6. American Nuclear Society meeting on fission-product behavior and source term research: proceedings. [PWR; BWR

    Microsoft Academic Search

    Huebner

    1985-01-01

    The Topical Meeting on Fission Product Behavior and Source Term Research, sponsored by the American Nuclear Society, was held at Snowbird, Utah, during July 15 to 19, 1984. The purpose was to bring together source term researchers, modelers of nuclear accidents, commercial nuclear power officials, and government regulators from all over the world to discuss source term issues and results

  7. Fission-product-behavior modeling in risk analysis: an assessment of the relevant phenomena. [PWR; BWR

    Microsoft Academic Search

    A. R. Taig; C. D. Leigh; D. A. Powers; J. L. Sprung; J. C. Cunnane; H. I. Avci; P. Baybutt; J. A. Gieseke; T. Margulies

    1983-01-01

    A review of the phenomenology governing the release and transport of fission products in LWR plants in severe accidents is described. Recommended approaches and models for incorporation into the MELCOR code for application in risk analysis are discussed. Major areas of phenomenological uncertainty and modeling difficulty are highlighted.

  8. THE MECHANISMS OF FISSION GAS DIFFUSION IN GRAPHITE. Summary of Work Completed for the Period June 1, 1962May 31, 1963

    Microsoft Academic Search

    W. S. Diethorn; P. L. Jr. Walker

    1963-01-01

    A study of fission gas diffusion in graphite including the diffusion ; mechanism and its relation to graphite structure and history is described. The ; parameters of crystallite size and perfection, pores, degree of graphitization, ; and radiation damage to the graphite lattice are of chief concern. The diffusion ; study involves the measurement of fission gas release from graphite

  9. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600C and 1700C, and at least 10 to 102 at 1800C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600C, 105 at 1700C, and 103 at 1800C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600C and 103 at 1700C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800C since all the compacts heated up at 1800C contain particles with failed SiC layers whose release dominates the release from intact particles. For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700C, and 102 to 103 at 1800C. These values might be somewhat over-estimated because the strontium retention during irradiation cannot be assessed a priori, which affects the magnitude of the calculated release during safety testing. The diffusivity of strontium in UCO cannot be derived from these heating tests, but it is assumed to be modeled correctly using the IAEA recommended value for kernel diffusivity. For krypton, there is no reliable release data for compacts heated up at 1600C, which includes all the compacts containing only intact particles. At 1700 and 1800C, comparisons show an over-prediction of the release from compacts containing particles with failed SiC by 1 to 1.5 orders of magnitude. The available data from these heating tests do not allow to determine which of the TRISO-coatings layers diffusivities are under or over-estimated.

  10. Release of fission products from fuel during the in-vessel phases of severe nuclear reactor accidents

    Microsoft Academic Search

    1983-01-01

    We discuss the phenomena involved in the release of volatile fission products during the in-vessel phases of a severe nuclear reactor accident. To date, much effort has been placed on the release of noble fission gases from solid fuel grains to grain surfaces and edges, the release of those gases from the fuel matrix, and the mechanical response of the

  11. Permissible overheating of the gaseous medium in a gas-flow laser excited by fission fragments from uranium nuclei

    SciTech Connect

    Prikhod'ko, E V; Sizov, A N [Russian Federal Nuclear Center 'All-Russian Scientific Research Institute of Experimental Physics', Sarov, Nizhnii Novgorod Region (Russian Federation)

    1999-09-30

    The limits on the permissible overheating of the gaseous mixtures in gas-flow lasers excited by uranium-nuclear-fission fragments are examined. The first limit is associated with the possibility of growth of a heat-removal zone near the walls and the second arises from the need to preserve the cavity stability. (active media)

  12. Nondestructive Evaluation of Transient Fission Gas Release from a Pulse-Irradiated PWR Segment Fuel by Counting Krypton 85

    Microsoft Academic Search

    Kazuaki YANAGISAWA

    1992-01-01

    An experimental study was made in the Nuclear Safety Research Reactor at JAERI to evaluate transient fission gas release (FGR) from high burn-up PWR type fuel rods using a nondestructive technique, the counting of radioactive Kr in the fuel plenum. Within the scope of this experiment, the following results were obtained:1. By using the pulse irradiation technique, the radioactivity of

  13. Analysis of fission gas release measurements using the COMETHE IIIJ and FCODE-Alpha computer codes. Final report. [PWR; BWR

    Microsoft Academic Search

    G. Leppert; L. Rayes; E. Rumble; R. Stuart

    1981-01-01

    Fission gas release predictions from FCODE-Alpha and COMETHE IIIJ were compared with experimental data from a representative group of light water reactor (LWR) fuel rods and with each other. In the first phase of the study, standard versions of the codes obtained from the Electric Power Software Center were compared with data from 36 rods. A modified version of COMETHE

  14. The potential to use fission gas release experiments to measure lattice and grain boundary diffusion in metallic fuels

    Microsoft Academic Search

    Wayne E. King; Martin Robel; George H. Gilmer

    2011-01-01

    We have applied a model for lattice and grain boundary diffusion in polycrystalline materials to assess the potential for the use of fission gas release experiments to measure the lattice and grain boundary diffusion coefficients in metallic nuclear fuel materials. Our assessment is that, assuming that grain boundary diffusion in metallic fuels is similar to that in other metals, it

  15. Yields of short-lived fission products produced following 235U(nth,f)

    NASA Astrophysics Data System (ADS)

    Tipnis, S. V.; Campbell, J. M.; Couchell, G. P.; Li, S.; Nguyen, H. V.; Pullen, D. J.; Schier, W. A.; Seabury, E. H.; England, T. R.

    1998-08-01

    Measurements of gamma-ray spectra, following the thermal neutron fission of 235U have been made using a high purity germanium detector at the University of Massachusetts Lowell (UML) Van de Graaff facility. The gamma spectra were measured at delay times ranging from 0.2 s to nearly 10 000 s following the rapid transfer of the fission fragments with a helium-jet system. On the basis of the known gamma transitions, forty isotopes have been identified and studied. By measuring the relative intensities of these transitions, the relative yields of the various precursor nuclides have been calculated. The results are compared with the recommended values listed in the ENDF/B-VI fission product data base (for the lifetimes and the relative yields) and those published in the Nuclear Data Sheets (for the beta branching ratios). This information is particularly useful for the cases of short-lived fission products with lifetimes of the order of fractions of a second or a few seconds. Independent yields of many of these isotopes have rather large uncertainties, some of which have been reduced by the present study.

  16. Modeling of fission gas effects observed in TREAT loss-of-flow test R8 using SAS3D. [LMFBR

    SciTech Connect

    Dunn, F.E.; Morris, E.E.

    1985-01-01

    The TREAT loss-of-flow test R8 has been analyzed using a modified version of the SAS3D accident analysis code in order to establish experimental verification of SAS3D modeling changes introduced to assess the effects of fission gas release from the upper gas plenum on coolant voiding dynamics and clad motion prior to the onset of fuel motion in whole core accident studies. The specific motivation for the analysis was the necessity of investigating the potential for fuel compaction by the fission gas stored in the gas plena at the upper end of the fuel pins in the Clinch River Breeder Reactor Project (CRBRP). These investigations were required to support licensing activities underway just prior to the demise of the project. In test R8, an unirradiated, seven-pin fuel bundle was subjected to a simulated Fast Flux Test Facility flow coastdown. The power was held constant at its nominal value.

  17. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    SciTech Connect

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. many mechamistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, reearch, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  18. Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

    Microsoft Academic Search

    Yang-Hyun Koo; Byung-Ho Lee; Jin-Sik Cheon; Dong-Seong Sohn

    2002-01-01

    To analyze the effect of an inhomogeneous mixture of an PuO2 powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO2 fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu

  19. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    SciTech Connect

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., /sup 137/Cs, /sup 90/Sr, /sup 129/I, /sup 99/Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,..cap alpha..), (n,..gamma..), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm/sup 3/ are required for a practical fusion-based fission product transmutation system.

  20. Massyield distributions of fission products from photo-fission of nat Pb induced by 5070MeV bremsstrahlung

    Microsoft Academic Search

    Haladhara Naik; Guinyun Kim; Ashok Goswami; Sarbjit Singh; Vijay Kumar Manchanda; Devesh Raj; Srinivasan Ganesan; Young Do Oh; Hee-Seock Lee; Kyung Sook Kim; Man-Woo Lee; Moo-Hyun Cho; In Soo Ko; Won Namkung

    2010-01-01

    The massyield distributions of various fission products have been determined in the 50-, 60- and 70-MeV end point bremsstrahlung\\u000a induced fission of natPb using off-line ?-ray spectrometric technique in the electron linac at Pohang Accelerator Laboratory, Korea. The massyield\\u000a distributions are symmetric with average mass of 102.34, 102.25 and 102.03 and FWHM of 21, 22 and 23 mass unit, respectively.

  1. FITPULS: a code for obtaining analytic fits to aggregate fission-product decay-energy spectra. [In FORTRAN

    SciTech Connect

    LaBauve, R.J.; George, D.C.; England, T.R.

    1980-03-01

    The operation and input to the FITPULS code, recently updated to utilize interactive graphics, are described. The code is designed to retrieve data from a library containing aggregate fine-group spectra (150 energy groups) from fission products, collapse the data to few groups (up to 25), and fit the resulting spectra along the cooling time axis with a linear combination of exponential functions. Also given in this report are useful results for aggregate gamma and beta spectra from the decay of fission products released from /sup 235/U irradiated with a pulse (10/sup -4/ s irradiation time) of thermal neutrons. These fits are given in 22 energy groups that are the first 22 groups of the LASL 25-group decay-energy group structure, and the data are expressed both as MeV per fission second and particles per fission second; these pulse functions are readily folded into finite fission histories. 65 figures, 11 tables.

  2. The measurement of compositions and the isotopic distribution of released fission gas in the fuel rods of pressurized water reactors (PWR) of Korea

    Microsoft Academic Search

    S. D. Park; D. K. Min; Y. K. Ha; K. Song

    2010-01-01

    In this work the analysis procedures of fission gas compositions and their isotopic distributions using a gas chromatography\\u000a (GC) system and\\/or a quadrupole mass spectrometer (QMS) system were established, and their analysis results were reviewed\\u000a in order to evaluate their analytical performance. Also, the accumulated data, up to now, regarding fission gas measurement\\u000a were reviewed to discern any irradiation histories

  3. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    SciTech Connect

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space.

  4. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    SciTech Connect

    Tang, I.N.; Munkelwitz, H.R.

    1982-07-01

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated.

  5. NEW ENDF/B-VII.0 EVALUATIONS OF NEUTRON CROSS SECTIONS FOR 32 FISSION PRODUCTS.

    SciTech Connect

    KIM,H.; LEE, Y.-O.; HERMAN, M.; MUGHABGHAB, S.F.; OBLOZINSKY, P.; ROCHMAN, D.

    2007-04-22

    Neutron cross sections for fission products play important role not only in the design of extended burnup core and fast reactors, but also in the study of the backend fuel cycle and the criticality analysis of spent fuel. New evaluations in both the resonance and fast neutron regions were performed by the KAERI-BNL collaboration for 32 fission products. These were {sup 95}Mo, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, and complete isotope chains of {sup 142-148,150}Nd, {sup 144,147,148-154}Sm, and {sup 156,158,160-164}Dy. The evaluations cover a large amount of reaction channels, including all those needed for neutronics calculations. Also, they cover the entire energy range, from 10{sup -5} eV to 20 MeV, including the thermal, resolved, and unresolved resonance regions, and the fast neutron region.

  6. High-power proton linac for transmuting the long-lived fission products in nuclear waste

    SciTech Connect

    Lawrence, G.P.

    1991-01-01

    High power proton linacs are being considered at Los Alamos as drivers for high-flux spallation neutron sources that can be used to transmute the troublesome long-lived fission products in defense nuclear waste. The transmutation scheme being studied provides a high flux (> 10{sup 16}/cm{sup 2}{minus}s) of thermal neutrons, which efficiently converts fission products to stable or short-lived isotopes. A medium-energy proton linac with an average beam power of about 110 MW can burn the accumulated Tc99 and I129 inventory at the DOE's Hanford Site within 30 years. Preliminary concepts for this machine are described. 3 refs., 5 figs., 2 tabs.

  7. High-temperature reactor fuel fission product release and distribution at 1600 to 1800 degrees C

    Microsoft Academic Search

    W. Schenk; H. Nabielek

    1991-01-01

    The essential feature of small, modular high-temperature reactors (HTRs) is the inherent limitation in maximum accident temperature to below 1600° C combined with the ability of coated particle fuel to retain all safety-relevant fission products under these conditions. To demonstrate this ability, spherical fuel elements with modern TRISO particles are irradiated and subjected to heating tests. Even after extended heating

  8. 216-S-1 and S-2 mixed-fission-product crib-characterization study

    SciTech Connect

    Van Luik, A. E.; Smith, R. M.

    1982-03-01

    The 216-S-1 and 2 crib is an underground structure that was used for the disposal of radioactively contaminated liquid waste at the Hanford Site. The crib received acidic, intermediate level, mixed fission-product waste solutions from 1952 to 1956. The 1980 status of radioactive contaminants in the sediment beneath the crib was investigated. The results indicate that the radionuclide distributions are stable, with no evidence of significant translocations found since the late 1960's.

  9. Continuous fission-product monitor system at Oyster Creek. Final report

    Microsoft Academic Search

    L. L. Collins; E. T. Chulick

    1980-01-01

    A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for

  10. Reconstruction of radial fission-product distributions in reactor fuels from a small number of projections

    SciTech Connect

    Barnes, B.K.; Phillips, J.R.; Barnes, M.L.

    1981-01-01

    Four mathematical techniques for reconstruction of the radial two-dimensional distribution of fission products using projections obtained by nondestructive gamma scanning were evaluated. Reconstruction of a picture from a finite set of projections is mathematically indeterminate; therefore, reconstruction techniques are heuristic, particularly when only a small number of projections are available. Of the techniques evaluated, the filtered backprojection algorithm provided the best reconstruction for simulated gamma-ray sources, as well as for actual irradiated fuel material.

  11. The separation of fission-product rare elements toward bridging the nuclear and soft energy systems

    Microsoft Academic Search

    Masaki Ozawa; Yoshihiko Shinoda; Yuichi Sano

    2002-01-01

    Based on the present state of the art of the separation technology, recycling of fission-product rare elements (FRE) in the FBR spent fuel is discussed. The rad.-waste fractionation is in accordance with the present society's trend toward zero-emission, and the mean of salt-free method utilizing electrochemistry agrees with the principles of the newly established green chemistry. A catalytic electrolytic extraction

  12. Exploratory study of fission product yields of neutron-induced fission of 235U , 238U , and 239Pu at 8.9 MeV

    NASA Astrophysics Data System (ADS)

    Bhatia, C.; Fallin, B. F.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E.; Bredeweg, T. A.; Fowler, M. M.; Moody, W.; Rundberg, R. S.; Rusev, G. Y.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2015-06-01

    Using dual-fission chambers each loaded with a thick (200 -400 -mg /c m2) actinide target of 235 ,238U or 239Pu and two thin (10 -100 -? g /c m2) reference foils of the same actinide, the cumulative yields of fission products ranging from 92Sr to 147Nd have been measured at En= 8.9 MeV . The 2H(d ,n ) 3He reaction provided the quasimonoenergetic neutron beam. The experimental setup and methods used to determine the fission product yield (FPY) are described, and results for typically eight high-yield fission products are presented. Our FPYs for 235U(n ,f ) , 238U(n ,f ) , and 239Pu(n ,f ) at 8.9 MeV are compared with the existing data below 8 MeV from Glendenin et al. [Phys. Rev. C 24, 2600 (1981), 10.1103/PhysRevC.24.2600], Nagy et al. [Phys. Rev. C 17, 163 (1978), 10.1103/PhysRevC.17.163], Gindler et al. [Phys. Rev. C 27, 2058 (1983), 10.1103/PhysRevC.27.2058], and those of Mac Innes et al. [Nucl. Data Sheets 112, 3135 (2011), 10.1016/j.nds.2011.11.009] and Laurec et al. [Nucl. Data Sheets 111, 2965 (2010), 10.1016/j.nds.2010.11.004] at 14.5 and 14.7 MeV, respectively. This comparison indicates a negative slope for the energy dependence of most fission product yields obtained from 235U and 239Pu , whereas for 238U the slope issue remains unsettled.

  13. Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization

    SciTech Connect

    Joshua R. Versey

    2013-08-01

    This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is based on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.

  14. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, chemisorption, and the decay heating of RCS structures and gases are adopted in the analysis. The RCS flow models based on the density-difference between the RCS and containment pedestal region are developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP is developed for the analysis. The REVAP is incorporated with the MARCH, TRAPMELT3 and NAUA codes of the Source Term Code Pack Package (STCP). The NAUA code is used to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors determining the magnitude of revaporization and subsequent release of the volatile fission product. 8 figs., 1 tab.

  15. Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

    PubMed

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In ? = ?? ((?Grxn(TC))/(RTC)) + ? were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ?Grxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores. PMID:25675358

  16. Fission Product Transport in TRISO Particle Layers under Operating and Off-Normal Conditions

    SciTech Connect

    Van der Ven, Anton; Was, Gary; Wang, Lumin; Taheri, Mitra

    2014-07-07

    The objective of this project is to determine the diffusivity and chemical behavior of key fission products (ag, Cs, I. Te, Eu and Sr) through SiC and PyC both thermally, under irradiation, and under stress using FP introduction techniques that avoid the pitfalls of past experiments. The experimental approach is to create thin PyC-SiC couples containing the fission product to be studied embedded in the PyC layer. These samples will then be subjected to high temperature exposures in a vacuum and also to irradiation at high temperature, and last, to irradiation under stress at high temperature. The PyC serves as a host layer, providing a means of placing the fission product close to the SiC without damaging the SiC layer by its introduction or losing the FP during heating. Experimental measurements of grain boundary structure and distribution (EBSD, HRTEM, APT) will be used in the modeling effort to determine the qualitative dependence of FP diffusion coefficients on grain boundary orientation, temperature and stress.

  17. Solvent Extraction of Plutonium(IV), Uranium(VI), and Some Fission Products with Di-n-octylsulfoxide

    Microsoft Academic Search

    J. P. Shukla; S. A. Pai; M. S. Subramanian

    1979-01-01

    Extraction behavior of plutonium(IV), uranium(VI), and some fission products from aqueous nitric acid media with di-n-octylsulfoxide (DOSO) has been studied over a wide range of conditions. Both the actinides are extracted essentially completely, whereas fission product contaminants like Zr, Ru, Ce, Eu, and Sr show negligible extraction. The absorption spectra of sulfoxide extracts containing either Pu or UO2 indicate the

  18. On-line ultrasonic gas entrainment monitor. [LMFBR fission gas bubble release

    Microsoft Academic Search

    C. K. Day; H. N. Pedersen

    1978-01-01

    Apparatus employing ultrasonic energy for detecting and measuring the quantity of gas bubbles present in liquids being transported through pipes is described. An ultrasonic transducer is positioned along the longitudinal axis of a fluid duct, oriented to transmit acoustic energy radially of the duct around the circumference of the enclosure walls. The back-reflected energy is received centrally of the duct

  19. Gamma-ray study of short-lived aggregate fission products from {sup 235}U(n,f)

    SciTech Connect

    Schier, W.A.; Campbell, J.M.; Couchell, G.P.; Li, S. [and others

    1993-10-01

    A high purity germanium detector was used to measure gamma-ray spectra following thermal neutron-induced fission of {sup 235}U for eighteen delay time intervals in the range 0.1-100,000 a following fission. A helium-jet system was used to rapidly transport fission products to a low-background counting area. The gamma-ray spectrometer employed beta-gamma coincidence as well as a Nal(Tl) annulus for background suppression. Nearly 300 gamma-ray peaks have been analyzed for delay-time intervals < 10 s after fission in the energy range 0-6 MeV. The time evolution of a peak provides information about the lifetime of the precursor nuclide as well as the ratio of its direct production in fission to its production through radioactive decay. Where decay schemes are known the measured gamma-ray intensities can be used to deduce relative yields and in some cases metastable-to-ground-state production probabilities. Results of lifetimes, yields, etc. of short-lived products will be compared with CINDER 10 calculations based on ENDF/B-IV fission-product data.

  20. Experimental determination of the antineutrino spectrum of the fission products of U238.

    PubMed

    Haag, N; Gtlein, A; Hofmann, M; Oberauer, L; Potzel, W; Schreckenbach, K; Wagner, F M

    2014-03-28

    An experiment was performed at the scientific neutron source FRM II in Garching to determine the cumulative antineutrino spectrum of the fission products of U238. Target foils of natural uranium were irradiated with a thermal and a fast neutron beam and the emitted ? spectra were recorded with a ?-suppressing electron telescope. The obtained ? spectrum of the fission products of U235 was normalized to the data of the magnetic spectrometer BILL. This method strongly reduces systematic errors in the U238 measurement. The ? spectrum of U238 was converted into the corresponding ?e spectrum. The final ?e spectrum is given in 250keV bins in the range from 2.875 to 7.625MeV with an energy-dependent error of 3.5% at 3MeV, 7.6% at 6MeV, and ?14% at energies ?7??MeV (68% confidence level). Furthermore, an energy-independent uncertainty of ?3.3% due to the absolute normalization is added. Compared to the generally used summation calculations, the obtained spectrum reveals a spectral distortion of ?10% but returns the same value for the mean cross section per fission for the inverse beta decay. PMID:24724646

  1. A cyclic time optimization approach to the study of 252Cf fission products

    NASA Astrophysics Data System (ADS)

    Price, R. I.; Ebong, I. D. U.; Adams, John A.; Roy, R. R.

    1980-05-01

    A K X-ray-beta particle coincidence technique has been investigated for the study of the beta decay of fission products from 252Cf. A fission-fragments transport system has been developed and its optimization curve used for the identification of the half-life associated with the K X-ray peak originating from the Mo ? Tc decay high-resolution lithium-drifted silicon spectrometer and a plastic scintillation spectrometer were used in the analysis of the K X-rays and beta particles respectively. A half-life of (0.98 0.03) min was associated with the K X-rays from technetium. A Kurie plot of the coincidence beta spectrum revealed at least three beta groups with end-point energies of (2.19 0.19) MeV, (1.64 0.14) MeV and (1.04 0.10) MeV.

  2. Experimental Decay Heat of Beta Particles from ^235U ^238U and ^239Pu Fission Products

    NASA Astrophysics Data System (ADS)

    Li, S.; Campbell, J. M.; Couchell, G. P.; Nguyen, H. V.; Pullen, D. J.; Seabury, E. H.; Schier, W. A.; Tipnis, S. V.; England, T.

    1996-10-01

    These results were obtained at the UMass Lowell 5.5 MV Van de Graaff accelerator and 1 MW research reactor. A He-jet/tape transport system was used to achieve delay times after fission as short as 0.4 s, where few experimental results exist. Measured beta spectra used a thin-disk-gating technique to reject accompanying gamma rays. Both beta and gamma sources were used in energy calibration. A set of trial responses for the beta spectrometer spanned electron energies 0-10 MeV. Spectra unfolded for energy distributions were compared with previous measurements. Measured beta count-rates using a pair of beta detectors provided relative normalization. Results of beta decay heat were compared to calculations based on ENDF/B-VI fission-product data. ^*Supported in part by the U.S. Department of Energy.

  3. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

    2009-05-05

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  4. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B. (Oakland, CA); Prussin, Stanley G. (Kensington, CA)

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  5. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    DOEpatents

    Norman, Eric B. (Oakland, CA); Prussin, Stanley G. (Kensington, CA)

    2009-01-06

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  6. Comments on fission-gas release from fuel at high burnup in Vol. 19, No. 6. [Water cooled reactors

    Microsoft Academic Search

    H. Ocken; J. T. A. Roberts

    1979-01-01

    Meyer, Beyer, and Voglewede have proposed that an enhancement factor be applied to existing vendor models when fission-gas release (FGR) at burnups greater than 20,000 MWd\\/metric ton is calculated for licensing purposes. This enhancement factor is derived from FGR data obtained from liquid-metal-cooled fast breeder reactor (LMFBR) fuel. The analysis assumes that the intrinsic source of the high FGR measured

  7. Fission product release and microstructure changes during laboratory annealing of a very high burn-up fuel specimen

    Microsoft Academic Search

    J.-P. Hiernaut; T. Wiss; J.-Y. Colle; H. Thiele; C. T. Walker; W. Goll; R. J. M. Konings

    2008-01-01

    A commercial PWR fuel sample with a local burn-up of about 240MWd\\/kgHM was annealed in a Knudsen cell mass spectrometer system with a heating rate of 10K\\/min up to 2750K at which temperature the sample was completely vaporized. The release of fission gases and fission products was studied as a function of temperature. In one of the runs the heating

  8. Partition of soluble fission products between the grey phase, ZrO2 and uranium dioxide

    NASA Astrophysics Data System (ADS)

    Cooper, M. W. D.; Middleburgh, S. C.; Grimes, R. W.

    2013-07-01

    The energies to remove fission products from UO2 or UO2+x and incorporate them into BaZrO3, SrZrO3 (grey phase constituent phases) and ZrO2 have been calculated using atomistic scale simulation. These energies provide the thermodynamic drive for partition of soluble fission products between UO2 or UO2+x and these secondary oxide constituents of the fuel system. Tetravalent cation partition into BaZrO3, SrZrO3 and ZrO2 was only preferable for species with smaller radii than Zr4+, regardless of uranium dioxide stoichiometry. Under stoichiometric conditions both the larger and the smaller trivalent cations were found to segregate to BaZrO3 but only the smaller fuel additive elements Cr3+ and Fe3+ segregate to SrZrO3. Partition from UO2+x was always unfavourable for trivalent cations. Additions of excess Cr3+ (as a fuel additive) are predicted make the partition into BaZrO3 and SrZrO3 more favourable from UO2 for the larger trivalent cations. Trivalent fission products with radii smaller than or equal to that of Sm3+ were identified to segregate into ZrO2 only from UO2. No segregation to SrO or BaO is predicted. Conventional Krger-Vink notation does not allow for distinction between oxygen sites in the UO2 and the secondary phases. As such, from now on we will distinguish all defects in the UO2 lattice with a line, e.g. MUׯ.

  9. Ion exchange in the atomic energy industry with particular reference to actinide and fission product separation

    SciTech Connect

    Jenkins, I.L.

    1984-01-01

    Reviewed are some of the uses of ion exchange processes used by the nuclear industry for the period April, 1978 to April, 1983. The topics dealt with are: thorium, protactinium, uranium, neptunium, plutonium, americium, cesium and actinide-lanthanide separations; the higher actinides - Cm, Bk, Cf, Es and Fm; fission products; ion exchange in the geological disposal of radioactive waste. Consideration is given to safety in the use of ion exchangers and in safe methods of disposal of such materials. Full scale and pilot plant process descriptions are included as well as summaries of laboratory studies. 130 references.

  10. Neutron economy and nuclear data for transmutation of long-lived fission products

    Microsoft Academic Search

    M. Igashira; T. Ohsaki

    2002-01-01

    In the study of Self-Consistent Nuclear Energy System, the following 29 long-lived fission products (LLFPs) have been selected to be transmuted into stable or short-lived nuclides: 106Ru, 102Rh, 109Cd, 125Sb, 134Cs, 146,147Pm, 154,155Eu, 171Tm, 85Kr, 90Sr, 93mNb, 113mCd, 121mSn, 137Cs, 151Sm, 152Eu, 108mAg, 158Tb, 166mHo, 79Se, 93Zr, 94Nb, 99Tc, 107Pd, 126Sn, 129I, 135Cs. In the present study, the number of

  11. Viability of long-lived fission products as signatures in forensic radiochemistry

    SciTech Connect

    McAninch, J.E.; Proctor, I.D.; Stoyer, N.J.; Moody, K.J.

    1997-01-01

    Forensic radiochemistry refers to studies on special nuclear materials, related to nonproliferation and anti-smuggling efforts. AMS (accelerator mass spectroscopy) measurement of long-lived fission products and U and Pu isotopes has the potential to significantly aid the field of forensic radiochemistry by providing new or more sensitive signatures and improving on the speed with which they can be determined. Expanding the suite of signatures obtainable form an illicit sample of special nuclear material increases the likelihood that its point of origin can be positively identified, leveraging LLNL`s impact on policy decisions regarding national security.

  12. Immobilization of fission products in low-temperature ceramic waste forms

    SciTech Connect

    Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Mandalika, V.

    1997-01-01

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics (CBPCs) for use in solidifying and stabilizing low-level mixed wastes. The focus of this work is development of CBPCs for use with fission-product wastes generated from high-level waste (HLW) tank cleaning or other decontamination and decommissioning activities. The volatile fission products such as Tc, Cs, and Sr removed from HLW need to be disposed of in a low-temperature immobilization system. Specifically, this paper reports on the solidification and stabilization of separated {sup 99}Tc from Los Alamos National Laboratory`s complexation-elution process. Using rhenium as a surrogate form technetium, we fabricated CBPC waste forms by acid-base reactions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with 35 wt.% waste loading. Standard leaching tests such as ANS 16.1 and PCT were conducted on the final waste forms. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water-immersion tests.

  13. Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (II)

    Microsoft Academic Search

    Y?ichir? KAMEMOTO; Koreyuki SHIBA; Muneo HANDA; Shigeru YAMAGISHI; Takeshi FUKUDA; Yoshihisa TAKAHASHI; Takaaki TANIFUJI; Shunz? ?MORI

    1967-01-01

    The ?-ray spectra of fission gases released from UO2-graphite pellets under neutron irradiation were measured. With and without separating fission gases into xenon and krypton, 25 kinds of ?-ray were observed and assigned to nine nuclides, Kr, Kr, Kr(Rb), Xe, Xe, Xe and Xe (Cs). A value of 15 min is proposed for the half-life of Xe, based on analysis

  14. Thermal Expansion of Simulated Fuels with Dissolved Fission Products in a UO2 Matrix

    NASA Astrophysics Data System (ADS)

    Kang, K. H.; Na, S. H.; Park, C. J.; Kim, Y. H.; Song, K. C.; Lee, S. H.; Kim, S. W.

    2009-06-01

    As a part of the DUPIC (direct use of spent PWR fuel in CANDU reactors) fuel development program, the thermal expansion of simulated spent fuel pellets with dissolved fission products has been studied by using a thermo-mechanical analyzer (TMA) in the temperature range from 298 K to 1773 K to investigate the effects of fission products forming solid solutions in a UO2 matrix on the thermal expansions. Simulated fuels with an equivalent burn-up of (30 to 120) GWd/tU were used in this study. The linear thermal expansions of the simulated fuel pellets were higher than that of UO2, and the difference between these fuel pellets and UO2 increased monotonically with temperature. For the temperature range from 298 K to 1773 K, the values of the average linear thermal expansion coefficients for UO2 and simulated fuels with an equivalent burn-up of (30, 60, and 120) GWd/tU are 1.19 10-5 K-1, 1.22 10-5 K-1, 1.26 10-5 K-1, and 1.32 10-5 K-1, respectively.

  15. High-level waste glass field burial test: leaching and migration of fission products

    SciTech Connect

    Melnyk, T.W.; Johnson, L.H.; Walton, F.B.

    1984-01-01

    In June 1960, 25 nepheline syenite-based glass hemispheres containing the fission products /sup 137/Cs, /sup 90/Sr, /sup 144/Ce and /sup 106/Ru were buried below the water table in a sandy-soil aquifer at the Chalk River Nuclear Laboratories of Atomic Energy of Canada Limited. Measurements of soil and groundwater concentrations of /sup 90/Sr and /sup 137/Cs have been interpreted using non-equilibrium migration models to deduce the leaching history of the glass for these burial conditions. The leaching history derived from the field data has been compared to laboratory leaching of samples taken from a glass hemisphere retrieved in 1978, and also to pre-burial laboratory leaching of identical hemispheres. The time dependence of the leach rates observed for the buried specimens suggests that leaching is inhibited by the formation of a protective surface layer. The effect of the kinetic limitations of the fission-product/sandy-soil interactions is discussed with respect to the migration of /sup 90/Sr and /sup 137/Cs over a 20 year time scale. It is concluded that kinetically limited sorption by oxyhdroxides, rather than equilibrium ion exchange, controls the long-term migration of /sup 90/Sr. Cesium is initially rapidly bound to the micaceous fraction of the sand, but slow remobilization of /sup 137/Cs in particulate form is observed and is believed to be related to bacterial action.

  16. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    SciTech Connect

    McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

    1996-10-01

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion.

  17. Fission Product Yields of {sup 233}U, {sup 235}U, {sup 238}U and {sup 239}Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    SciTech Connect

    Laurec, J.; Adam, A.; Bruyne, T. de [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Bauge, E., E-mail: eric.bauge@cea.f [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G. [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Authier, N.; Casoli, P. [Commissariat a l'Energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2010-12-15

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for {sup 235}U(n,f), {sup 239}Pu(n,f) in a thermal spectrum, for {sup 233}U(n,f), {sup 235}U(n,f), and {sup 239}Pu(n,f) reactions in a fission neutron spectrum, and for {sup 233}U(n,f), {sup 235}U(n,f), {sup 238}U(n,f), and {sup 239}Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  18. The Outlook for Some Fission Products Utilization with the Aim to Immobilize Long-Lived Radionuclides

    SciTech Connect

    Pokhitonov, Y.A. [Khlopin Radium Institute, St. Petersburg (Russian Federation)

    2008-07-01

    The prospects for development of nuclear power are intimately associated with solving the problem of safe management and removal from the biosphere of generated radioactive wastes. The most suitable material for fission products and actinides immobilization is the crystalline ceramics. By now numerous literature data are available concerning the synthesis of a large range of various materials with zirconium-based products. It worth mentioning that zirconium is only one of fission products accumulated in the fuel in large amounts. The development of new materials intended for HLW immobilization will allow increasing of radionuclides concentration in solidified product so providing costs reduction at the stage of subsequent storage. At the same time the idea to use for synthesis of compounds, suitable as materials for long-term storage or final disposal of rad-wastes some fission products occurring in spent fuel in considerable amount and capable to form insoluble substances seems to be rather attractive. In authors opinion in the nearest future one can expect the occurrence of publications proposing the techniques allowing the use of 'reactor's zirconium, molybdenum or, perhaps, technetium as well, with the aim of preparing materials suitable for long-lived radionuclides storage or final disposal. The other element, which is generated in the reactor and worth mentioning, is palladium. The prospects for using palladium are defined not only by its higher generation in the reactor, but by a number of its chemical properties as well. It is evident that the use of natural palladium with the purpose of radionuclides immobilization is impossible due to its high cost and deficiency). In author's opinion such materials could be used as targets for long-lived radionuclides transmutation as well. The object of present work was the study on methods that could allow to use 'reactor' palladium with the aim of long-lived radionuclides such as I-129 and TUE immobilization. In the paper the results of experiments on synthesis of matrices with TUE oxides and PdI{sub 2} on palladium base are presented. (authors)

  19. Electron Microscopic Evaluation and Fission Product Identification of Irradiated TRISO Coated Particles from the AGR-1 Experiment: A Preliminary Review

    SciTech Connect

    IJ van Rooyen; DE Janney; BD Miller; PA DEmkowicz; J Riesterer

    2014-05-01

    Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this paper a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objectives of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. Microstructural characterization focused on fission-product precipitates in the SiC-IPyC interface, the SiC layer and the fuel-buffer interlayer. The results provide significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates with significantly higher concentrations of Pd and U. Different approaches to resolving this problem are discussed. An initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations were observed and no debonding of the SiC-IPyC interlayer as a result of irradiation was observed for the samples investigated. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  20. Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary Study

    SciTech Connect

    I J van Rooyen; D E Janney; B D Miller; J L Riesterer; P A Demkowicz

    2012-10-01

    ABSTRACT Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this presentation a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objective of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. The characterization emphasized fission-product precipitates in the SiC-IPyC interface, SiC layer and the fuel-buffer interlayer, and provided significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentration Ag in precipitates with significantly higher concentrations of contain Pd and U. Different approaches to resolving this problem are discussed. Possible microstructural differences between particles with high and low releases of Ag particles are also briefly discussed, and an initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations or debonding of the SiC-IPyC interlayer as a result of irradiation were observed. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  1. Progress in Chile in the development of the fission {sup 99}Mo production using modified CINTICHEM

    SciTech Connect

    Schrader, R.; Klein, J.; Medel, J.; Marin, J.; Salazar, N.; Barrera, M.; Albornoz, C.; Chandia, M.; Errazu, X.; Becerra, R.; Sylvester, G.; Jimenez, J.C. [Chilean Nuclear Energy Commission, CCHEN, Amunategui 95, Santiago (Chile); Vargas, E. [Mechanical Engineering Faculty, Pontificia Universidad Catolica de Valparaiso, Valparaiso (Chile)

    2008-07-15

    Fission {sup 99}Mo will be produced in Chile irradiating low-enriched uranium (LEU) foil in a MTR research reactor. For the purpose of developing the capability to fabricate the target, which is done of uranium foil enclosed in swaged concentric aluminum tubes, dummy targets are being fabricated using 130 {mu}m copper foil instead of the uranium foil, wrapped in a 14{mu}m nickel fission-recoil barrier. Dummy targets using several dimensions of copper foil have been assembled; however, the emphasis is being set in targets fabricated using the dimensions of the LEU foil that KAERI will provide, i.e. 50 mm x 100mm x 0.130 mm. The assembling of target using the last dimensions has not been free of difficulties. Neutronic calculations and preliminary thermal and fluid analyses were performed to estimate the fission products activity and the heat removal capability for a 13 grams LEU-foil annular target, which will be irradiated in the RECH-1 research reactor at the level power of 5 MW during 48 hours. In a fume hood, Cintichem processing of natural uranium shavings with the addition of different carriers were performed, obtaining recovery over 90% of the added Mo carrier. Expertise has been gained in (a) foil dissolution process in a dissolver locally designed, (b) in Mo precipitation process, and (c) preparation of the purification columns with AgC, C and HZrO. Additionally, the irradiated target cutting machine with an innovative design was finally assembled. (author)

  2. Ab initio molecular dynamics of high-temperature unimolecular dissociation of gas-phase RDX and its dissociation products.

    PubMed

    Schweigert, Igor V

    2015-03-26

    Unimolecular dynamics of gas-phase hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX) and its dissociation products were simulated using density functional theory (DFT) at the M06-L level. The simulations of RDX at 2000 K showed that dissociation proceeds from multiple conformers, mostly via homolytic fission of an N-N bond with a minor contribution from elimination of HONO, in agreement with previous transition state theory calculations. However, the simulations of the fission and elimination products revealed that secondary N-N fission is facile and, at the simulated temperature of 1750 K, dominant over other mechanisms. The simulations of the resulting intermediates revealed a number of new unimolecular pathways that have not been previously considered. The transition structures and minimal energy paths were calculated for all reactions to confirm these observations. Based on these findings, a revised set of the unimolecular reactions contributing to gas-phase RDX decomposition is proposed. PMID:25738393

  3. Sorption of {sup 239}Np and {sup 235}U fission products by zeolite Y, Mexican natural erionite, and bentonite

    SciTech Connect

    Olguin, M.T.; Solache, M.; Iturbe, J.L. [Instituto Nacional de Investigaciones Nucleares, C.P. (Mexico)]|[Universidad Autonoma Metropolitana, C.P. (Mexico)] [and others

    1996-09-01

    Zeolite Y, erionite, and bentonite have been used in this work to remove {sup 239}Np and {sup 235}U fission products from aqueous solutions at various pH values. It was found that the sorption of fission products by aluminosilicates takes place by different mechanisms, mainly ion exchange, precipitation, and electrostatic surface interaction. The radionuclides content was determined by {gamma}-spectrometry, and X-ray diffraction was used to learn whether the solids maintained their crystallinity at different pH values.

  4. Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

    SciTech Connect

    Besmann, Theodore M [ORNL] [ORNL; Ferber, Mattison K [ORNL] [ORNL; Lin, Hua-Tay [ORNL] [ORNL

    2014-01-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  5. CINDER; M0102; point depletion fission product. [CDC6600; IBM360; FORTRAN IV (CDC6600), FORTRAN IV and BAL (IBM360)

    Microsoft Academic Search

    2008-01-01

    CINDER is a four-group, one-point depletion and fission product program based on the evaluation of a general analytical solution of nuclides coupled in any linear sequence of radioactive decays and neutron absorptions in a specified neutron flux spectrum. The desired depletion and fission product chains and all physical data are specified by the problem originator. The program computes individual nuclide

  6. Use of Gamma spectrometry for measuring fission product releases during a simulated PWR severe accident: Application to the VERDON experimental program

    Microsoft Academic Search

    G. Ducros; S. Bernard; M. P. Ferroud-Plattet; O. Ichim

    2009-01-01

    The release of fission products (FP) from a pressurized water reactor (PWR) during a hypothetical severe accident is a major topic in nuclear reactor safety assessment, since they are the main contributors to the source term in the environment. Fission products with short half lives are of particular importance due to their potential high radiological effects. In order to precisely

  7. Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

    NASA Astrophysics Data System (ADS)

    Mushtaq, A.; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab

    2009-04-01

    Low enriched uranium foil (19.99% 235U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/ 99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.

  8. Investigation of the Distribution of Fission Products Silver, Palladium and Cadmium in Neutron Irradiated SIC using a Cs Corrected HRTEM

    SciTech Connect

    I. J. van Rooyen; E. Olivier; J. H Neethlin

    2014-10-01

    Electron microscopy examinations of selected coated particles from the first advanced gas reactor experiment (AGR-1) at Idaho National Laboratory (INL) provided important information on fission product distribution and chemical composition. Furthermore, recent research using STEM analysis led to the discovery of Ag at SiC grain boundaries and triple junctions. As these Ag precipitates were nano-sized, high resolution transmission electron microscopy (HRTEM) examination was used to provide more information at the atomic level. This paper describes some of the first HRTEM results obtained by examining a particle from Compact 4-1-1, which was irradiated to an average burnup of 19.26% fissions per initial metal atom (FIMA), a time average, volume-averaged temperature of 1072C; a time average, peak temperature of 1182C and an average fast fluence of 4.13 x 1021 n/cm2. Based on gamma analysis, it is estimated that this particle may have released as much as 10% of its available Ag-110m inventory during irradiation. The HRTEM investigation focused on Ag, Pd, Cd and U due to the interest in Ag transport mechanisms and possible correlation with Pd, Ag and U previously found. Additionally, Compact 4-1-1 contains fuel particles fabricated with a different fuel carrier gas composition and lower deposition temperatures for the SiC layer relative to the Baseline fabrication conditions, which are expected to reduce the concentration of SiC defects resulting from uranium dispersion. Pd, Ag, and Cd were found to co-exist in some of the SiC grain boundaries and triple junctions whilst U was found to be present in the micron-sized precipitates as well as separately in selected areas at grain boundaries. This study confirmed the presence of Pd both at inter- and intragranular positions; in the latter case specifically at stacking faults. Small Pd nodules were observed at a distance of about 6.5 micron from the inner PyC/SiC interface.

  9. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  10. Capture of volatile iodine, a gaseous fission product, by zeolitic imidazolate framework-8.

    PubMed

    Sava, Dorina F; Rodriguez, Mark A; Chapman, Karena W; Chupas, Peter J; Greathouse, Jeffery A; Crozier, Paul S; Nenoff, Tina M

    2011-08-17

    Here we present detailed structural evidence of captured molecular iodine (I(2)), a volatile gaseous fission product, within the metal-organic framework ZIF-8 [zeolitic imidazolate framework-8 or Zn(2-methylimidazolate)(2)]. There is worldwide interest in the effective capture and storage of radioiodine, as it is both produced from nuclear fuel reprocessing and also commonly released in nuclear reactor accidents. Insights from multiple complementary experimental and computational probes were combined to locate I(2) molecules crystallographically inside the sodalite cages of ZIF-8 and to understand the capture of I(2) via bonding with the framework. These structural tools included high-resolution synchrotron powder X-ray diffraction, pair distribution function analysis, and molecular modeling simulations. Additional tests indicated that extruded ZIF-8 pellets perform on par with ZIF-8 powder and are industrially suitable for I(2) capture. PMID:21766858

  11. LOW-FIDELITY CROSS SECTION COVARIANCES FOR 219 FISSION PRODUCTS IN THE FIRST NEUTRON REGION.

    SciTech Connect

    PIGNI,M.T.; HERMAN, M.; OBLOZINSKY, P.; ROCHMAN, D.

    2007-04-27

    An extensive set of covariances for neutron cross sections in the energy range 5 keV-20 MeV has been developed to provide initial, low-fidelity but consistent uncertainty data for nuclear criticality safety applications. The methodology for the determination of such covariances combines the nuclear reaction model code EMPIRE, which calculates sensitivity to nuclear reaction model parameters, and the Bayesian code KALMAN to propagate uncertainty of the model parameters to cross sections. Taking into account the large scale of the project (219 fission products), only partial reference to experimental data has been made. Therefore, the covariances are, to a large extent, derived from the perturbation of several critical model parameters selected through the sensitivity analysis. These parameters define optical potential, level densities and pre-equilibrium emission. This work represents the first attempt ever to generate nuclear data covariances on such a scale.

  12. A new half-life measurement of the long-lived fission product 126Sn

    NASA Astrophysics Data System (ADS)

    Haas, P.; Gartenmann, P.; Golser, R.; Kutschera, W.; Suter, M.; Synal, H.-A.; Wagner, M. J. M.; Wild, E.; Winkler, G.

    1996-06-01

    The half-life of the long-lived fission product 126Sn has been determined through a specific activity measurement to be (2.07 0.21) 10 5 a. The measurement was performed with 126Sn material extracted from spent fuel rods of a nuclear power reactor. The activity concentration in this material was measured to be 4.97 0.15 Bq {126Sn }/{mg Sn>}. The half-life was determined by combining this activity concentration with the isotopic abundance of {126Sn }/{Sn} = (9.23 0.87) 10 -6. [P. Gartenmann et al., this issue, preceeding paper]. The latter was measured by accelerator mass spectrometry (AMS) and is reported in an adjacent paper. This is the first direct measurement of the half-life of 126Sn, which previously had been estimated to be 10 5 years.

  13. Evaluation of six decontamination processes on actinide and fission product contamination

    SciTech Connect

    Conner, C.; Chamberlain, D.B.; Chen, L. [Argonne National Lab., IL (United States)] [and others

    1995-12-31

    In-situ decontamination technologies were evaluated for their ability to: (1) reduce equipment contamination levels to allow either free release of the equipment or land disposal, (2) minimize residues generated by decontamination, and (3) generate residues that are compatible with existing disposal technologies. Six decontamination processes were selected. tested and compared to 4M nitric acid, a traditional decontamination agent: fluoroboric acid (HBF{sub 4}), nitric plus hydrofluoric acid, alkaline persulfate followed by citric acid plus oxalic acid, silver(II) plus sodium persulfate plus nitric acid, oxalic acid plus hydrogen peroxide plus hydrofluoric acid, and electropolishing using nitric acid electrolyte. The effectiveness of these solutions was tested using prepared 304 stainless steel couponds contaminated with uranium, plutonium, americium, or fission products. The decontamination factor for each of the solutions and tests conditions were determined; the results of these experiments are presented.

  14. Cherenkov light detection as a velocity selector for uranium fission products at intermediate energies

    NASA Astrophysics Data System (ADS)

    Yamaguchi, T.; Enomoto, A.; Kouno, J.; Yamaki, S.; Matsunaga, S.; Suzaki, F.; Suzuki, T.; Abe, Y.; Nagae, D.; Okada, S.; Ozawa, A.; Saito, Y.; Sawahata, K.; Kitagawa, A.; Sato, S.

    2014-12-01

    The in-flight particle separation capability of intermediate-energy radioactive ion (RI) beams produced at a fragment separator can be improved with the Cherenkov light detection technique. The cone angle of Cherenkov light emission varies as a function of beam velocity. This can be exploited as a velocity selector for secondary beams. Using heavy ion beams available at the HIMAC synchrotron facility, the Cherenkov light angular distribution was measured for several thin radiators with high refractive indices (n = 1.9 ~ 2.1). A velocity resolution of ~10-3 was achieved for a 56Fe beam with an energy of 500 MeV/nucleon. Combined with the conventional rigidity selection technique coupled with energy-loss analysis, the present method will enable the efficient selection of an exotic species from huge amounts of various nuclides, such as uranium fission products at the BigRIPS fragment separator located at the RI Beam Factory.

  15. The effect of lattice and grain boundary diffusion on the redistribution of Xe in metallic nuclear fuels: Implications for the use of ion implantation to study fission-gas-bubble nucleation mechanisms

    Microsoft Academic Search

    Wayne E. King; Scott J. Tumey; Jeffrey Rest; George H. Gilmer

    2011-01-01

    A multi-atom gas bubble-nucleation mechanism has been proposed as part of a predictive fission-gas release model for metallic nuclear fuels. Validation of this mechanism requires experimental measurement of fission-gas bubble-size distributions at well-controlled gas concentrations and temperatures. There are advantages to carrying out such a study using ion implantation as the source of gas atoms compared with neutron irradiations. In

  16. Apparatus for the dynamic and total measurement of retained fission gas

    Microsoft Academic Search

    J. W. Early; R. M. Abernathey

    1982-01-01

    This versatile apparatus provides a quick, accurate and inexpensive determination of fission gases Kr and Xe in irradiated nuclear fuels. Samples are heated to 2000°C in a vacuum furnace under controlled temperature-time conditions and the released Kr and Xe are dynamically and integrally measured by a quadrupole mass spectrometer.

  17. Laser microsampling method for determination of retained fission gas in irradiated nuclear fuels. [LMFBR

    Microsoft Academic Search

    D. G. Graczyk; G. Bandyopadhyay; S. M. Gehl; J. P. Hughes; H. T. Goodspeed

    1979-01-01

    A small ruby laser adapted to fire through a microscope is used to release fission gases from specific sites on a plane surface of an irradiated fuel specimen. Interaction of the focused laser pulse with the specimen surface results in a conical crater from which sampled material has been vaporized; the crater is surrounded by a heat-affected zone in which

  18. Actinide Recovery Experiments with Bench-Scale Liquid Cadmium Cathode in Fission Product-Laden Molten Salt

    Microsoft Academic Search

    S. X. Li; S. D. Herrmann; R. W. Benedict; K. M. Goff; M. F. Simpson

    2009-01-01

    This article summarizes the observations and analytical results from a series of bench- scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+\\/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the LCC experiments. Actinide recovery

  19. Thermodynamics of fission products in dispersion fuel designs - first principles modeling of defect behavior in bulk and at interfaces

    SciTech Connect

    Liu, Xiang-yand [Los Alamos National Laboratory; Uberuaga, Blas P [Los Alamos National Laboratory; Nerikar, Pankaj [Los Alamos National Laboratory; Sickafus, Kurt E [Los Alamos National Laboratory; Stanek, Chris R [Los Alamos National Laboratory

    2009-01-01

    Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO{sub 2} and UO{sub 2} oxides, and the MgO/(U, Hf, Ce)O{sub 2} interfaces have been carried out. In the case of UO{sub 2}, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-sit Coulomb repulsion between the localized 5f electrons. The fission product solution energies in bulk UO{sub 2{+-}x} have been calculated as a function of non-stoichiometry x, and were compared to that in MgO. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO{sub 2}. However, this trend is reversed or reduced for alkaline earth oxides with larger cation sizes. The solution energies of fission products in MgO are substantially higher than in UO{sub 2{+-}x}, except for the case of Sr in the hypostoichiometric case. Due to size effects, the thermodynamic driving force of segregation for Xe and Cs from bulk MgO to the MgO/fluorite interface is strong. However, this driving force is relatively weak for Sr.

  20. Fission product deposition behavior during the in-pile severe fuel damage test SFD 1-3

    Microsoft Academic Search

    K. Vinjamuri; D. H. Meikrantz; J. D. Baker; D. J. Osetek

    1986-01-01

    Retention mechanisms of radioactive material in the reactor coolant system are important because they have a significant effect on the consequences of a severe nuclear power plant accident. Fission products are normally retained in the fuel; however, during a severe accident, a significant quantity of radioactive gases and some aerosols may be expected to escape from the fuel region. Vapor

  1. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    Microsoft Academic Search

    James C. Marra; Amanda Y. Billings; Jarrod V. Crum; Joseph V. Ryan; John D. Vienna

    2010-01-01

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product

  2. Fission product plateout and liftoff in the MHTGR primary system: A review

    SciTech Connect

    Wichner, R.P. (Oak Ridge National Lab., TN (USA))

    1991-04-01

    A review is presented of the technical basis for predicting radioactivity release resulting from depressurization of an MHTGR primary system. Consideration is restricted to so called dry events with no involvement of the steam system. The various types of deposition mechanisms effective for iodine, cesium, strontium, and silver are discussed in terms of their chemical characteristics and the nature of the materials in the primary system. Emphasis is given to iodine behavior, including means for estimating the quantity available for release, the types of plateout locations in the primary system, and the effect of dust on distribution and release. The behavior of fission products cesium, strontium, and silver in such accidents is presented qualitatively. A major part of the review deals with expected dust levels, types, and transport. Available information on the level and nature of dust in the HTGR primary system is reviewed. A summary is presented of dust deposition and liftoff mechanisms. It was concluded that recent approaches to dust liftoff modeling, based on turbulent burst concepts for removal from surfaces, probably offer advantages over the current shear ratio approach. This study concludes that iodine releases from dry depressurization events are likely to be extremely low, on the order of millicuries, due to a predictably low degree of chemical desorption, a low degree of dust liftoff, and a low involvement of iodine with dust. It was also concluded that deposition mechanisms controlling the distribution of fission product material in the primary system, and hence also controlling the degree of liftoff, depend strongly on the chemical nature of the individual elements. Therefore contrary to the current practice, both plateout and liftoff models should reflect those unique chemical and physical properties. 56 refs., 16 figs., 23 tabs.

  3. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    NASA Astrophysics Data System (ADS)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than 7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  4. Production Trends of Shale Gas Wells

    E-print Network

    Khan, Waqar A.

    2010-01-14

    Station, TX, May (1998). 3. El-Banbi, A.H. and Wattenbarger, R.A.: ?Analysis of Linear Flow in Gas Well Production,? paper SPE 39972 presented at the 1998 SPE Gas Technology Symposium, Calgary, 15-18 March. 4. Ar?valo-Villagr?n, J...

  5. ConocoPhillips Gas Hydrate Production Test

    SciTech Connect

    Schoderbek, David; Farrell, Helen; Howard, James; Raterman, Kevin; Silpngarmlert, Suntichai; Martin, Kenneth; Smith, Bruce; Klein, Perry

    2013-06-30

    Work began on the ConocoPhillips Gas Hydrates Production Test (DOE award number DE-NT0006553) on October 1, 2008. This final report summarizes the entire project from January 1, 2011 to June 30, 2013.

  6. Analysis of gas production methods for methane gas hydrate reservoirs

    NASA Astrophysics Data System (ADS)

    Ivakhnenko, Aleksandr; Baluanov, Bakhytzhan; Shopenova, Aigerim; Gulnur, Asan; Agzomova, Bagdagul

    2015-04-01

    In methane gas hydrate reservoir (MH), pressure and temperature conditions are in the MH stability region in the initial stage. To dissociate MH and produce gas from a MH reservoir, pressure and temperature conditions should be moved to the dissociation region. Therefore, three methods of depressurization, thermal and inhibitor injection have been modeled and analyzed as a basic methods for different conditions that might occur in nature. Furthermore, several methods such as injection of gas other than methane and irradiation of ultrasonic wave were also investigated especially for the MH dissociation and possible gas production. The simulation results allowed to select optimal screening approach for the appropriate production method that can be employed in specific MH conditions.

  7. Grain boundary sweeping and liquefaction-induced fission product behavior in nuclear fuel under severe-core damage accident conditions

    Microsoft Academic Search

    Rest

    1984-01-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from: (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests performed at Oak Ridge National Laboratory; and (2) trace-irradiated and high-burnup LWR fuel during severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory

  8. RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS

    SciTech Connect

    Clark, E.

    2013-08-31

    Data from previous reports on studies of polymers exposed to tritium gas is further analyzed to estimate rates of radiolytic gas production. Also, graphs of gas release during tritium exposure from ultrahigh molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, a trade name is Teflon), and Vespel polyimide are re-plotted as moles of gas as a function of time, which is consistent with a later study of tritium effects on various formulations of the elastomer ethylene-propylene-diene monomer (EPDM). These gas production rate estimates may be useful while considering using these polymers in tritium processing systems. These rates are valid at least for the longest exposure times for each material, two years for UHMW-PE, PTFE, and Vespel, and fourteen months for filled and unfilled EPDM. Note that the production rate for Vespel is a quantity of H{sub 2} produced during a single exposure to tritium, independent of length of time. The larger production rate per unit mass for unfilled EPDM results from the lack of filler- the carbon black in filled EPDM does not produce H{sub 2} or HT. This is one aspect of how inert fillers reduce the effects of ionizing radiation on polymers.

  9. MELCOR 1.8.5 modeling aspects of fission product release, transport and deposition an assessment with recommendations.

    SciTech Connect

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels. This paper discusses the synthesis of these findings in the MELCOR severe accident code. Based on recent assessments of MELCOR 1.8.5 fission product release modeling against the Phebus FPT-1 test and on observations from the ISP-46 exercise, modifications to the default MELCOR 1.8.5 release models are recommended. The assessments identified an alternative set of Booth diffusion parameters recommended by ORNL (ORNL-Booth), which produced significantly improved release predictions for cesium and other fission product groups. Some adjustments to the scaling factors in the ORNL-Booth model were made for selected fission product groups, including UO{sub 2}, Mo and Ru in order to obtain better comparisons with the FPT-1 data. The adjusted model, referred to as 'Modified ORNL-Booth,' was subsequently compared to original ORNL VI fission product release experiments and to more recently performed French VERCORS tests, and the comparisons was as favorable or better than the original CORSOR-M MELCOR default release model. These modified ORNL-Booth parameters, input to MELCOR 1.8.5 as 'sensitivity coefficients' (i.e. user input that over-rides the code defaults) are recommended for the interim period until improved release models can be implemented into MELCOR. For the case of ruthenium release in air-oxidizing conditions, some additional modifications to the Ru class vapor pressure are recommended based on estimates of the RuO{sub 2} vapor pressure over mildly hyperstoichiometric UO{sub 2}. The increased vapor pressure for this class significantly increases the net transport of Ru from the fuel to the gas stream. A formal model is needed. Deposition patterns in the Phebus FPT-1 circuit were also significantly improved by using the modified ORNL-Booth parameters, where retention of lower volatile Cs{sub 2}MoO{sub 4} is now predicted in the heated exit regions of the FPT-1 test, bringing down depositions in the FPT-1 steam generator tube to be in closer alignment with the experimental data. This improvement in 'RCS' deposition behavior preserves the overall correct release of cesium to the containment that was observed even with the default CORSOR-M model. Not correctly treated however is the release and transport of Ag to the FPT-1 containment. A model for Ag release from control rods is presently not available in MELCOR. Lack of this model is thought to be responsible for the underprediction by a factor of two of the total aerosol mass to the FPT-1 containment. It is suggested that this underprediction of airborne mass led to an underprediction of the aerosol agglomeration rate. Underprediction of the agglomeration rate leads to low predictions of the aerosol particle size in comparison to experimentally measured ones. Small particle size leads low predictions of the gravitational settling rate relative to the experimental data. This error, however, is a conservative one in that too-low settling rate would result in a larger source term to the environment. Implementation of an interim Ag release model is currently under study. In the course of this assessment, a review of MELCOR release models was performed and led to the identification of several areas for future improvements to MELCOR. These include upgrading the Booth release model to account for changes in local oxidizing/reducing conditions and including a fuel oxidation model to accommodate effects of fuel stoichiometry. Models such as implemented in the French ELSA code and described by Lewis are considered appropriate for MELCOR. A model for ruthenium release under air oxidizing conditions is also needed and should be included as part of a fuel oxidation model since fuel stoichiometry is a fundamen

  10. Helium and fission gas behaviour in magnesium aluminate spinel and zirconia for actinide transmutation

    Microsoft Academic Search

    P. M. G. Damen

    2003-01-01

    In order to reduce the long-term radiotoxicity of spent nuclear fuel, many studies are performed on partitioning and transmutation of actinides. In such a scenario, the long-lived radio-isotopes (mostly actinides) are partitioned from the nuclear waste, and subsequently transmuted or fissioned in a neutron flux in shorter-lived or stable isotopes. In order to place the actinides in a neutron flux,

  11. Determination of long-lived fission products and actinides in Savannah River site HLW sludge and glass for waste acceptance

    SciTech Connect

    Bibler, N.E.; Boyce, W.T.; Coleman, C.J. [and others

    1997-10-01

    Savannah River Site (SRS) is currently immobilizing the radioactive, caustic, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository acceptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma - mass spectrometry and {alpha}, {beta}, and {gamma} counting methods. Examples of the radionuclides are Sr-90, Cs-137, U-238, Pu-239, and Cm-244. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradioactive glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic area in proportion to their yields from the fission of U-235 in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass.

  12. Yields of short-lived fission products produced following {sup 235}U(n{sub th},f)

    SciTech Connect

    Tipnis, S.V.; Campbell, J.M.; Couchell, G.P.; Li, S.; Nguyen, H.V.; Pullen, D.J.; Schier, W.A.; Seabury, E.H. [University of Massachusetts Lowell, Lowell, Massachusetts 01854 (United States)] [University of Massachusetts Lowell, Lowell, Massachusetts 01854 (United States); England, T.R. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)] [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    1998-08-01

    Measurements of gamma-ray spectra, following the thermal neutron fission of {sup 235}U have been made using a high purity germanium detector at the University of Massachusetts Lowell (UML) Van de Graaff facility. The gamma spectra were measured at delay times ranging from 0.2 s to nearly 10thinsp000 s following the rapid transfer of the fission fragments with a helium-jet system. On the basis of the known gamma transitions, forty isotopes have been identified and studied. By measuring the relative intensities of these transitions, the relative yields of the various precursor nuclides have been calculated. The results are compared with the recommended values listed in the ENDF/B-VI fission product data base (for the lifetimes and the relative yields) and those published in the Nuclear Data Sheets (for the beta branching ratios). This information is particularly useful for the cases of short-lived fission products with lifetimes of the order of fractions of a second or a few seconds. Independent yields of many of these isotopes have rather large uncertainties, some of which have been reduced by the present study. {copyright} {ital 1998} {ital The American Physical Society}

  13. Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

    SciTech Connect

    Wright, A.L. [Oak Ridge National Lab., TN (United States)

    1994-06-01

    This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena and presents major conclusions on the state of the art.

  14. Fission product release and fuel behavior of irradiated light water reactor fuel under severe accident conditions. The ACRR ST-1 Experiment

    SciTech Connect

    Allen, M.D.; Stockman, H.W.; Reil, K.O. [Sandia National Labs., Albuquerque, NM (United States); Fisk, J.W. [Tills (Jack) and Associates, Inc., Albuquerque, NM (United States)

    1991-11-01

    The annular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain time-resolved data on the release of fission products from irradiated fuels under well-controlled light water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST-1 was conducted in a highly reducing environment at a system pressure of approximately 0.19 MPa, and at maximum fuel temperatures of about 2490 K. The data will be used for the development and validation of mechanistic fission product release computer codes such as VICTORIA.

  15. Deposition and distribution of Chernobyl fallout fission products and actinides in a Russian soil profile.

    PubMed

    Carbol, P; Solatie, D; Erdmann, N; Nyln, T; Betti, M

    2003-01-01

    In this article the distribution of fission products and actinides in a soil profile from Novo Bobovicky in Russia, which was contaminated due to the Chernobyl nuclear power plant accident, is described. The ground deposition of long-lived fission products determined by gamma-spectrometry was (recalculated to 26 April 1986) 1600 kBq (137)Cs/m(2), 900 kBq (134)Cs/m(2) and 60 kBq (125)Sb/m(2). Of these radionuclides (137)Cs shows the dominating activity at the present time. After 6.5 years 90% of the Cs and Sb activity was contained in the upper 4 cm. A (239,240)Pu ground deposition of 77.4+/-8.0 Bq/m(2) was determined by alpha-spectrometry. The (238)Pu/(239,240)Pu activity ratio of 0.30+/-0.03 and (241)Pu/(239,240)Pu activity ratio of 115+/-14 (in 1986) measured in the soil profile, indicates that the analysed Pu originates mainly from the Chernobyl accident. The average (234)U/(238)U activity ratio of 1.06+/-0.29 indicates that the uranium in this soil is dominated by naturally occurring uranium. The alpha- and beta-autoradiography revealed that the activity is mainly present in particulate form. It could further be observed that the spots containing alpha- or beta-activity originated from different particles. A comparison of alpha-autoradiography with the bulk Pu and Am activity showed that 92% of the alpha-activity was present as clearly detectable alpha-spots. The beta-active particles, located by beta-autoradiography were correlated with gamma-spectrometric measurements and contained only (137)Cs. These hot spots ranged from 0.02 to 0.15 Bq.It could be concluded that the vertical transport of (137)Cs and fuel fragments occurs mainly by movement of particles through the soil. It could also be concluded that the fuel fragments found, in this soil were depleted in respect to Cs, Sb and Eu. Comparison of the analysed (238)Pu/(239,240)Pu, (241)Pu/(239,240)Pu and (241)Am/(239,240)Pu ratios with the ratios calculated with ORIGEN-S code gave an estimate of the average burn-up of the fuel particles to be in the range of 11-12 GWd/tU. The results presented in this article are valid for this single soil profile and should not be generalised unless validated in a more rigorous study of a larger number of soil profiles. PMID:12726697

  16. Measurement of fission gas release, internal pressure and cladding creep rate in the fuel pins of PHWR bundle of normal discharge burnup

    Microsoft Academic Search

    U. K. Viswanathan; D. N. Sah; B. N. Rath; S. Anantharaman

    2009-01-01

    Fuel pins of a Pressurised Heavy Water Reactor (PHWR) fuel bundle discharged from Narora Atomic Power Station unit #1 after attaining a fuel burnup of 7528MWd\\/tU have been subjected to two types of studies, namely (i) puncture test to estimate extent of fission gas release and internal pressure in the fuel pin and (ii) localized heating of the irradiated fuel

  17. Rare gas studies in Luna 16-G-7 fines by stepwise heating technique - A low fission solar wind Xe.

    NASA Technical Reports Server (NTRS)

    Kaiser, W. A.

    1972-01-01

    Examination of He, Ne, Kr, and Xe in a dust sample (equal to or less than 125 micrometer) of Luna 16 in 12 temperature steps with especially small intervals in the low temperature range (80 C steps). The gas concentrations, as well as their relative abundances, are in general agreement with values reported by Vinogradov (1971) for Luna 16 and values found in Apollo 11 fines except for Ne. Comparison is made with various other experimental results. The solar wind Xe was lower in the fission-affected isotopes than was found in Apollo 11 fines and in the 1000 C fraction of the Pesyanoe meteorite as measured by Marti (1969). Air-Xe is interpreted as a fractional solar wind Xe with the composition found in this study.

  18. Methane hydrate gas production by thermal stimulation

    SciTech Connect

    McGuire, P.L.

    1981-01-01

    Two models have been developed to bracket the expected gas production from a methane hydrate reservoir. The frontal-sweep model represents the upper bound on the gas production, and the fracture-flow model represents the lower bound. Parametric studies were made to determine the importance of a number of variables, including porosity, bed thickness, injection temperature, and fracture length. These studies indicate that the hydrate-filled porosity should be at least 15%, reservoir thickness should be about 25 ft or more, and well spacing should be fairly large (maybe 40 acres/well), if possible. Injection temperatures should probably be between 150 and 250/sup 0/F to achieve an acceptable balance between high heat losses and unrealistically high injection rates. Numerous important questions about hydrate gas production remain unanswered.

  19. Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima Dai-ichi reactor accident

    E-print Network

    S. MacMullin; G. K. Giovanetti; M. P. Green; R. Henning; R. Holmes; K. Vorren; J. F. Wilkerson

    2012-10-03

    We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^3 and 0.42 +/- 0.07 mBq/m^3 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  20. Measurement of Airborne Fission Products in Chapel Hill, NC, USA from the Kukushima Dai-ichi Reactor Accident

    SciTech Connect

    MacMullin, S. [Univ, of North Carolina & Triangle Universities Nucl. Lab - Durham, NC; Giovanetti, G. K. [Univ, of North Carolina & Triangle Universities Nucl. Lab - Durham, NC; Green, M. P. [Univ, of North Carolina & Triangle Universities Nucl. Lab - Durham, NC; Henning, R. [Univ, of North Carolina & Triangle Universities Nucl. Lab - Durham, NC; Holmes, R. [Univ. North Carolina-Chapel & Univ. of Illinois-Urbana; Vorren, K. [University of North Carolina / Triangle Universities Nuclear Lababoratory, Durham; Wilkerson, J. F. [UNC/Triangle Univ. Nucl. Lab, Durham, NC/ORNL

    2012-01-01

    We present measurement results of airborne fission products in Chapel Hill, NC, USA, from 62 d following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products 131I and 137Cs were measured with maximum activity concentrations of 4.2 0.6 mBq/m3 and 0.42 0.07 mBq/m3 respectively. Additional activity from 131,132I, 134,136,137Cs and 132Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

  1. Fast-neutron interaction with the fission product {sup 103}Rh

    SciTech Connect

    Smith, A.B. [Argonne National Lab., IL (United States)]|[Arizona Univ., Tucson, AZ (United States); Guenther, P.T. [Argonne National Lab., IL (United States)

    1993-09-01

    Neutron total and differential elastic- and inelastic-scattering cross sections of {sup 103}Rh are measured from {approximately} 0.7 to 4.5 MeV (totals) and from {approximately} 1.5 to 10 MeV (scattering) with sufficient detail to define the energy-averaged behavior of the neutron processes. Neutrons corresponding to excitations of groups of levels at 334 {plus_minus} 13, 536 {plus_minus} 10, 648 {plus_minus} 25, 796 {plus_minus} 20, 864 {plus_minus} 22, 1120 {plus_minus} 22, 1279 {plus_minus} 60, 1481 {plus_minus} 27 and 1683 {plus_minus} 39 keV were observed. Additional groups at 1840 {plus_minus} 79 and 1991 {plus_minus} 71 key were tentatively identified. Assuming the target is a collective nucleus reasonably approximated by a simple one-phonon vibrator, spherical-optical, dispersive-optical, and coupled-channels models were developed from the data base with attention to the parameterization of the large inelastic-scattering cross sections. The physical properties of these models are compared with theoretical predictions and the systematics of similar model parameterizations in this mass region. In particular, it is shown that the inelastic-scattering cross section of the {sup 103}Rh fission product is large at the relatively low energies of applied interest.

  2. Monitoring methods and dose assessment for internal exposures involving mixed fission and activation products containing actinides.

    PubMed

    Thind, K S

    2001-01-01

    Internal dose assessment for intakes of radionuclide mixtures is a difficult task. When the radionuclide mixture contains both the easy to detect gamma emitters, e.g., 60Co and 95Zr, and difficult to detect alpha emitters such as 239Pu and 241Am, a single monitoring method, such as in-vivo counting, is inadequate for detection and dose assessment. Recent experience with task related monitoring for such radionuclide mixtures at Ontario Power Generation CANDU nuclear power plants has offered an opportunity to review this topic and suggest a strategy for monitoring that involves a combination of in-vivo and in-vitro methods. Using the radionuclide composition data in a mixture from an actual case as an example, this paper describes a monitoring strategy for mixed fission and activation products, including the advantages and pitfalls of reliance on surrogate radionuclides for signaling the presence of actinides in the mixture. The described monitoring strategy is consistent with the recommendations of ICRP Publication 78, which advocates a "combination of techniques so as to make the best possible evaluation of an unusual situation, for example, a programme of both body activity and excreta measurements." The use of experience and professional judgement for interpreting the combined in-vivo and in-vitro data for interim and ultimate intake and dose assessment is discussed and emphasized. PMID:11204117

  3. Disposal of type-II long-lived fission products into outer space

    SciTech Connect

    Takahashi, Hiroshi; Chen, Xinyi

    1996-12-31

    The authors propose an alternative approach to dispose of long-lived fission products (LLFPs) of type-II, such as {sup 79}Se, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, {sup 135}Cs, and long-lived radioactive {sup 93}Zr into outer solar space. An escape velocity from the solar system of 42 km/s will be provided from either a parking orbit or the moon`s surface using an electrostatic accelerator and by neutralizing the charged accelerated LLFPs ions. LLFP ions must be neutralized to avoid their being trapped in earth and solar magnetic fields; almost 100% neutralization can be achieved by recirculating the non-neutralized ions through a magnetic field in the neutralizing device. This mode of disposition requires 2.2 kW power to eject most of the LLFPs generated by one LWR. This process is much smaller than a medium-energy proton beam power, a few tens of MW, which would be necessary to transmute these LLFPs using spallation neutrons created by protons. Due to their low radioactivity composed of mainly beta decay and low-energy gamma-rays, the shielding needed is not excessive and can be easily accommodated.

  4. Disposal of long-lived fission products into the outer solar system

    SciTech Connect

    Takahashi, Hiroshi; Chen Xinyi; Yu An [Brookhaven National Laboratory, Upton, New York, 11973 (United States)

    2002-07-01

    We propose approach to dispose of Long-Lived Fission Products (LLFPs) of type II such as {sup 99}Tc and {sup 129}I into outer solar space by providing an escape velocity from the solar system of 42 km/sec from a parking orbit or the moon's surface using a electrostatic accelerator and neutralizing the charged ions. LLFPs disposed uniformly in outer solar space pose no hazard as do LLFPs packages in Earth orbit, and have no effects on astronomical observations. This mode of disposition requires energy in the order of 1 keV for each nucleus, which is far smaller than the propulsion energy needed for launching a LLFPs package by rocket. Further, the power required of an accelerator ejecting most of the LLFPs generated by one LWR is 2.2 kW, which is much smaller than a medium-energy proton accelerator, a few tens of MW, which would be necessary to transmute these LLFPs using spallation neutrons created by protons. Ion thrusters, which has been developed for maneuvering rocket, might be used for disposition of LLFP instead of the a static accelerator, its usability is discussed. (authors)

  5. Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios

    E-print Network

    Alajo, Ayodeji Babatunde

    2011-08-08

    Studies of Texas A&M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY May 2010 Major Subject: Nuclear Engineering FISSION PRODUCT IMPACT REDUCTION VIA PROTRACTED IN...-CORE RETENTION IN VERY HIGH TEMPERATURE REACTOR (VHTR) TRANSMUTATION SCENARIOS A Dissertation by AYODEJI BABATUNDE ALAJO Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements...

  6. Use of an ions thruster to dispose of type II long-lived fission products into outer space

    SciTech Connect

    Takahashi, H.; Yu, A.

    1997-04-01

    To dispose of long-lived fission products (LLFPs) into outer space, an ions thruster can be used instead of a static accelerator. The specifications of the ions thrusters which are presently studies for space propulsion are presented, and their usability discussed. Using of a rocket with an ions thruster for disposing of the LLFPs directly into the sun required a larger amount of energy than does the use of an accelerator.

  7. Observation and Measurement of 79Se in Savannah River Site High Level Waste Tank Fission Product Waste

    Microsoft Academic Search

    R. A. Dewberry; J. D. Leyba; W. T. Boyce

    2000-01-01

    We report the first observation of confirmed 79Se activity in Savannah River Site high level fission product waste. 79Se was measured after a seven step chemical treatment to remove interfering activity from 137Cs, 90Sr, and plutonium at levels 105 times higher than the observed 79Se content and to remove 99Tc at levels 300 times higher than observed 79Se. 79Se was

  8. Delayed fission product gamma-ray transmission through low enriched uranium dioxide fuel pin lattices in air

    Microsoft Academic Search

    Timothy H. Trumbull

    2004-01-01

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray

  9. Analysis of the MIT research reactor fission product and actinide radioactivity inventories

    E-print Network

    Kennedy, William B. (William Blake), 1979-

    2004-01-01

    The current analysis of the MITR core radioactivity inventory eliminates unnecessary assumptions made in previous estimates of the inventory, and revises the list of contributory isotopes to include all actinide and fission ...

  10. 8 Breckenridge, D.G. et al. (2003) Caspase cleavage product of BAP31 induces mitochondrial fission through endoplasmic reticulum calcium

    E-print Network

    Baas, Peter W.

    2003-01-01

    mitochondrial fission machinery. J. Biol. Chem. 278, 3637336379 18 Tondera, D. et al. (2004) Knockdown of MTP18 product viral mitochondrion- localized inhibitor of apoptosis. J. Virol. 77, 631641 21 Poncet, D. et al

  11. Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

    NASA Astrophysics Data System (ADS)

    Honaiser, Eduardo Henrique Rangel

    The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

  12. An assessment of the radiological doses resulting from accidental uranium aerosol releases and fission product releases from a postulated criticality accident at the Oak Ridge Y-12 Plant

    SciTech Connect

    Fisher, S.E.; Lenox, K.E.

    1995-03-01

    A dose assessment for two separate normalized source terms was conducted for the Oak Ridge Y-12 Plant. The first source term consisted of the noble gas and iodine fission products emanating from a postulated criticality with a magnitude of 10{sup 19} fissions. The second postulated source term was 1 kg of respirable highly enriched uranium. The MELCOR Accident Consequence Code System 2 (MACCS2) (beta test) computer code was used for this assessment. Both fixed weather (e.g., constant weather assumed throughout the accident) and sampled weather cases were performed using MACCS2. The results of the analysis are summarized in terms of the effective dose equivalent as a function of distance along the downwind centerline of the plume. In addition, population doses for the workers and the public are presented. A brief code comparison between the MACCS2 and MESORAD computer codes is also presented. Modeling differences for the cloudshine and groundshine dose pathways are discussed. Finally, the dose results are summarized, and recommendations are provided that enable the reader to make quick estimates of downwind doses for different source terms that are scalable.

  13. EXTRACTION OF Am FROM NITRIC ACID BY CARBAMOYL-PHOSPHORYL EXTRACTANTS: THE INFLUENCE OF SUBSTITUENTS ON THE SELECTIVITY OF Am OVER Fe AND SELECTED FISSION PRODUCTS

    Microsoft Academic Search

    E. Philip Horwltz; Kathleen A. Martin; Herbert Diamond; Louis Kaplan

    1986-01-01

    A number of neutral extractants containing the P(0)(CH2)nC(0)N raolety were evaluated for their ability to extract Am from nitric acid and their selectivity for Am over Fe and selected fission products. Extractants containing alkoxy, alkyl, and aryl substltuents were evaluated. Tetrachloroethylene was used as a diluent. Fission products selected for study were Y, Zr, Mo, Tc, Ru, Rh, Pd, La,

  14. Uranium dioxide films with xenon filled bubbles for fission gas behavior studies

    NASA Astrophysics Data System (ADS)

    Usov, I. O.; Dickerson, R. M.; Dickerson, P. O.; Byler, D. D.; McClellan, K. J.

    2014-09-01

    Electron beam evaporation and ion beam assisted deposition (IBAD) methods were utilized to fabricate depleted UO2 films and UO2 films with embedded Xe atoms, respectively. The films were fabricated at elevated temperature of 700 C and also subsequently annealed at 1000 C to induce grain growth and Xe atom redistribution. The goal of this work was to synthesize reference UO2 samples with controlled microstructures and Xe-filled bubble morphologies, without the effects attendant to rector irradiation-induced fission. Transmission electron microscopy (TEM) microstructural characterization revealed that fine Xe-filled bubbles nucleated in the as grown films and subsequent annealing resulted in noticeable bubble size increase. Reported results demonstrate the great potential IBAD techniques and UO2 films have for various areas of nuclear materials studies.

  15. Interconnecting compressors control coalbed gas production

    SciTech Connect

    Payton, R.; Niederhofer, J. (Taurus Exploration Inc., Tuscaloosa, AL (United States))

    1992-10-05

    This paper reports that centralized compressors afford Taurus Exploration Inc.'s coalbed gas operations optimum control of gas production. Unlike satellite stations, the centralized system allows methane gas to e shifted from station to station via the interconnecting low-pressure pipeline network. The operations area encompasses approximately 40,000 acres, about 40 miles southwest of Birmingham, Ala. The project includes about 250-miles of low-pressure gas flow lines to almost 400 wells. The centralized system is less costly than a satellite station to build and operate. Unlike a satellite station that requires each compressor to have a complete set of ancillary equipment, the centralized system requires only one suction manifold, one dehydration setup, and one metering facility for every five compressor sets.

  16. Bio-gas production from alligator weeds

    NASA Technical Reports Server (NTRS)

    Latif, A.

    1976-01-01

    Laboratory experiments were conducted to study the effect of temperature, sample preparation, reducing agents, light intensity and pH of the media, on bio-gas and methane production from the microbial anaerobic decomposition of alligator weeds (Alternanthera philoxeroides. Efforts were also made for the isolation and characterization of the methanogenic bacteria.

  17. Synthesis Gas Production from Partial Oxidation of Methane with Air in AC Electric Gas Discharge

    E-print Network

    Mallinson, Richard

    Synthesis Gas Production from Partial Oxidation of Methane with Air in AC Electric Gas Discharge K 73019 Received October 11, 2002 In this study, synthesis gas production in an AC electric gas discharge power used in each condition, since the maximum methane and oxygen conversions and synthesis gas

  18. Actinide Recovery Experiments with Bench-Scale Liquid Cadmium Cathode in Fission Product-Laden Molten Salt

    SciTech Connect

    S. X. Li; S. D. Herrmann; R. W. Benedict; K. M. Goff; M. F. Simpson

    2009-02-01

    This article summarizes the observations and analytical results from a series of bench- scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the LCC experiments. Actinide recovery efficiency and Pu/U ratio in the liquid cadmium cathode product under variable conditions are reported in the article. Separation factors for actinides and rare earth elements in the salt/cadmium system are also presented.

  19. Characteristics of gas dynamics of flow lasers excited by fission fragments of uranium nuclei

    SciTech Connect

    Borovkov, V.V.; Lazhintsev, B.V.; Sizov, A.N. [Institute of Experimental Physics, Nizhny Novgorod Province (Russian Federation)] [and others] [Institute of Experimental Physics, Nizhny Novgorod Province (Russian Federation); and others

    1995-12-01

    The conceptual design of a nuclear-pumped cw laser is put forward. Alternation of laser cells with plane uranium layers and heat exchangers (radiators) in a shared gas loop can reduce the gas velocity down to {approximately} 10 m s{sup {minus}1}. The results are reported of an investigation of optical inhomogeneities which appear in He and Ar due to excitation of the active medium in a prototype flow laser. It is shown that, in a section perpendicular to the plane of the uranium layers, a pumping inhomogeneity creates a positive parabolic gas lens and in a section parallel to these layers an optical gas wedge is formed. A vortex zone is observed in the gas flow at the exits from heat exchangers. Simulation experiments demonstrate that this effect increases tens of times the thermal diffusivity of the gas and results in considerable refractive losses of radiation in the effective heat-exchange region. Methods of compensating for optical inhomogeneities and for reducing the influence of vortices are proposed. 17 refs., 5 figs.

  20. Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

    NASA Astrophysics Data System (ADS)

    Perry, R. J.; Dean, C. J.; Nichols, A. L.

    2014-06-01

    The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library.

  1. An innovative acoustic sensor for first in-pile fission gas release determination REMORA 3 experiment

    Microsoft Academic Search

    E. Rosenkrantz; J. Y. Ferrandis; F. Augereau; T. Lambert; D. Fourmentel; X. Tiratay

    2011-01-01

    A fuel rod has been instrumented with a new design of an acoustic resonator used to measure in a non destructive way the internal rod plenum gas mixture composition. This ultrasonic sensor has demonstrated its ability to operate in pile during REMORA 3 irradiation experiment carried out in the OSIRIS Material Testing Reactor (CEA Saclay, France). Due to very severe

  2. Fission product release from a pressurized water reactor defective fuel rod: effect of thermal cycling

    Microsoft Academic Search

    G. Kurka; A. Harrer; P. Chenebault

    1979-01-01

    The emission of fission gases and iodines by a pressurized water reactor fuel rod containing a defect when it is initially put in the reactor is studied experimentally using a pressurized water loop in the Siloe reactor at Grenoble. The initial leakage is simulated by making a small hole near the upper end of the rod. The rare gases and

  3. Feasibility of a digester gas fuel production facility

    SciTech Connect

    Dakes, G.; Greene, D.S.; Sheehan, J.F.

    1982-03-01

    Results of studies on the feasibility of using digester gas produced from wastewater sludge to fuel vehicles are reported. Availability and suitability of digester gas as well as digester gas production records and test analyses on digester gas were reviewed. The feasibility of the project based on economic and environmental considerations is reported and compared to possible alternative uses of the digester gas.

  4. Effects of Oil and Gas Production on Groundwater

    Microsoft Academic Search

    John James Tintera; Leslie Savage

    Section 91.101 of the Texas Natural Resources Code provides the RRC with jurisdiction over the full scope of oil and gas exploration, development, and production operations and activities, including the drilling of wells associated with oil and gas activities, gas plants, natural gas or natural gas liquids processing plants, pressure maintenance plants, underground hydrocarbon storage facilities, and activities associated with

  5. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2x: Implications for nuclear fuel performance modeling

    NASA Astrophysics Data System (ADS)

    Andersson, D. A.; Garcia, P.; Liu, X.-Y.; Pastore, G.; Tonks, M.; Millett, P.; Dorado, B.; Gaston, D. R.; Andrs, D.; Williamson, R. L.; Martineau, R. C.; Uberuaga, B. P.; Stanek, C. R.

    2014-08-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x non-stoichiometry were used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Ris fuel rod irradiation experiment was simulated.

  6. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

    SciTech Connect

    Giovanni Pastore; Michael R. Tonks; Derek R. Gaston; Richard L. Williamson; David Andrs; Richard Martineau

    2014-03-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Ris fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

  7. The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

    Microsoft Academic Search

    1995-01-01

    The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental

  8. Measurement and analysis of fission gas release from BNFL's SBR MOX fuel

    Microsoft Academic Search

    R. J White; S. B Fisher; P. M. A Cook; R Stratton; C. T Walker; I. D Palmer

    2001-01-01

    Puncture results are presented for seven SBR MOX fuel rods from the first prototypical commercial irradiation that was carried out in the Beznau-1 PWR. The rod average burn-up ranged from 31.2 to 35.6 MWd\\/kgHM. Comparison is made with the percentage of gas released from French MOX fuels and UO2 fuel. The results show that in the burn-up range investigated, SBR

  9. Extraction of plutonium(IV), uranium(VI) and some fission products by di-n-hexyl sulphoxide

    Microsoft Academic Search

    S. A. Pai; J. P. Shukla; P. K. Khopkar; M. S. Subramanian

    1978-01-01

    The extraction of nitric acid, plutonium, uranium and fission products such as zirconium, ruthenium and europium has been\\u000a investigated using di-n-hexyl sulphoxide in Solvesso-100. Results indicate that Pu(IV), U(VI), Zr(IV) and Ru NO(III) are extracted\\u000a as disolvates, whereas Eu(III) is extracted as the trisolvate. The absorption spectra of the plutonium(IV) and uranium(VI)\\u000a complexes extracted are similar to those of the

  10. Impact of Zr metal and coking reactions on the fission product aerosol release during MCCI (Molten Core Concrete Interactions)

    SciTech Connect

    Lee, M.; Davis, R.E.; Khatib-Rahbar, M.

    1987-01-01

    During a core meltdown accident in a light water reactor, molten core materials (corium) could leave the reactor vessel and interact with concrete. In this paper, the impact of the zirconium content of the corium pool and the coking reaction on the release of fission products during Molten Core Concrete Interactions (MCCI) are quantified using CORCON/MOD2 and VANESA computer codes. Detailed calculations show that the total aerosol generation is proportional to the zirconium content of the corium pool. Among the twelve fission product groups treated by the VANESA code, CsI, CsO/sub 2/ and Nb/sub 2/O/sub 5/ are completely released over the course of the core/concrete interaction, while an insignificant quantity of Mo, Ru and ZrO/sub 2/ are predicted to be released. The release of BaO, SrO and CeO/sub 2/ increase with increased Zr content, while the releases of Te and La/sub 2/O/sub 3/ are relatively unaffected by the Zr content of the corium pool. The impact of the coking reaction on the radiological releases is estimated to be significant; while the impact of the coking reaction on the aerosol production is insignificant.

  11. Natural gas hydrates - issues for gas production and geomechanical stability

    E-print Network

    Grover, Tarun

    2008-10-10

    Natural gas hydrates are solid crystalline substances found in the subsurface. Since gas hydrates are stable at low temperatures and moderate pressures, gas hydrates are found either near the surface in arctic regions or in deep water marine...

  12. Corrosion update: Oil and gas production

    SciTech Connect

    Treseder, R.S.; Tuttle, R.N.

    1995-12-31

    This 4 volume revised edition includes the chemical compositions and trade names for corrosion resistant alloys. This is the corrosion resource for designers, builders, and operators of oil and gas facilities. The book provides a comprehensive review of the available corrosion literature in oil and gas production. More than 1,200 papers are included, presented at various meetings and conferences, both US and international. In addition to listing the article titles and sources, the editors go beyond a typical abstract listing by providing the scope, conclusions, and method for ordering each paper. The computer disk is TextWare-based, permitting suers to conduct more extensive searches quickly, as well as providing the ability to print the search results. Includes four volumes and one computer floppy disk.

  13. Bio gas oil production from waste lard.

    PubMed

    Hancsk, Jeno; Baladincz, Pter; Kasza, Tams; Kovcs, Sndor; Tth, Csaba; Varga, Zoltn

    2011-01-01

    Besides the second generations bio fuels, one of the most promising products is the bio gas oil, which is a high iso-paraffin containing fuel, which could be produced by the catalytic hydrogenation of different triglycerides. To broaden the feedstock of the bio gas oil the catalytic hydrogenation of waste lard over sulphided NiMo/Al(2)O(3) catalyst, and as the second step, the isomerization of the produced normal paraffin rich mixture (intermediate product) over Pt/SAPO-11 catalyst was investigated. It was found that both the hydrogenation and the decarboxylation/decarbonylation oxygen removing reactions took place but their ratio depended on the process parameters (T = 280-380C, P = 20-80 bar, LHSV = 0.75-3.0? h(-1) and H(2)/lard ratio: 600 ?Nm(3)/m(3)). In case of the isomerization at the favourable process parameters (T = 360-370C, P = 40-50 bar, LHSV = 1.0? h(-1) and H(2)/hydrocarbon ratio: 400? Nm(3)/m(3)) mainly mono-branching isoparaffins were obtained. The obtained products are excellent Diesel fuel blending components, which are practically free of heteroatoms. PMID:21403875

  14. Bio Gas Oil Production from Waste Lard

    PubMed Central

    Hancsk, Jen?; Baladincz, Pter; Kasza, Tams; Kovcs, Sndor; Tth, Csaba; Varga, Zoltn

    2011-01-01

    Besides the second generations bio fuels, one of the most promising products is the bio gas oil, which is a high iso-paraffin containing fuel, which could be produced by the catalytic hydrogenation of different triglycerides. To broaden the feedstock of the bio gas oil the catalytic hydrogenation of waste lard over sulphided NiMo/Al2O3 catalyst, and as the second step, the isomerization of the produced normal paraffin rich mixture (intermediate product) over Pt/SAPO-11 catalyst was investigated. It was found that both the hydrogenation and the decarboxylation/decarbonylation oxygen removing reactions took place but their ratio depended on the process parameters (T = 280380C, P = 2080 bar, LHSV = 0.753.0?h?1 and H2/lard ratio: 600?Nm3/m3). In case of the isomerization at the favourable process parameters (T = 360370C, P = 40 50 bar, LHSV = 1.0?h?1 and H2/hydrocarbon ratio: 400?Nm3/m3) mainly mono-branching isoparaffins were obtained. The obtained products are excellent Diesel fuel blending components, which are practically free of heteroatoms. PMID:21403875

  15. Recovery of fission product rare earth sulfates from Purex 1WW

    Microsoft Academic Search

    E. J. Wheelwright; W. H. Swift

    1961-01-01

    Cerium-144 and 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°C and centrifugation at 90°C to take advantage of the lower solubility of the double sulfate salt at a higher temperature. The

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

    Microsoft Academic Search

    E. J. Wheelwright; W. H. Swift

    1961-01-01

    Cerium- and 144 promethium-147, accompanied by rare earths resulting ;\\u000a from fission or decay can be removed from Purex 1WW in>90% yield as an ;\\u000a insoluble, crystalline sodium-rare earth double sulfate. Precipitation is ;\\u000a initiated by a one-to-three hour equilibration at 90 deg C and centrifugation at ;\\u000a 90 deg C to take advantage of the lower solubility of the

  17. Cross Sections and Yields for the Photo-Fission Productions of {sup 238}U

    SciTech Connect

    Badamsambuu, J.; Khuukhenkhuu, G.; Norov, N.; Zuzaan, P. [Nuclear Research Center, National University of Mongolia, Ulaanbaatar (Mongolia); Belov, A. G.; Gangrsky, Yu. P. [Flerov Laboratory of Nuclear Reactions, JINR, Dubna (Russian Federation)

    2009-03-31

    The yields and reaction cross-sections of {sup 92}Sr, {sup 97}Zr, {sup 97}Nb and {sup 135}I at the photofission of {sup 238}U were measured. These fission-fragments have some peculiarities in nuclear structure or in practical using. The measurements were performed on the bremsstrahlung of FLNR JINR microtron, in the electron energy range 10-22 MeV. The activation method with Ge(Li)--detector was used in these measurements.

  18. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  19. Nuclear Fission

    NASA Astrophysics Data System (ADS)

    Denschlag, J. O.

    This chapter first gives a survey on the history of the discovery of nuclear fission. It briefly presents the liquid-drop and shell models and their application to the fission process. The most important quantities accessible to experimental determination such as mass yields, nuclear charge distribution, prompt neutron emission, kinetic energy distribution, ternary fragment yields, angular distributions, and properties of fission isomers are presented as well as the instrumentation and techniques used for their measurement. The contribution concentrates on the fundamental aspects of nuclear fission. The practical aspects of nuclear fission are discussed in http://dx.doi.org/10.1007/978-1-4419-0720-2_57 of Vol. 6.

  20. Measurements of Methane Emissions at Natural Gas Production Sites

    E-print Network

    Lightsey, Glenn

    Measurements of Methane Emissions at Natural Gas Production Sites in the United States #12;Why = 21 #12;Need for Study Estimates of methane emissions from natural gas production , from academic in assumptions in estimating emissions Measured data for some sources of methane emissions during natural gas

  1. Understanding the Basics of Gas Exploration and Production

    NSDL National Science Digital Library

    Albert, Eric K.

    This presentation from Eric K. Albert explains the basics of gas exploration and production, as well as some of the career opportunities created by the industry. Most of the presentation focuses on natural gas development, exploration and production. He also discusses where the jobs are in the natural gas industry.The presentation may be downloaded in Power Point file format.

  2. Microstructural characterization of irradiated U-7Mo/Al-5Si dispersion fuel to high fission density

    NASA Astrophysics Data System (ADS)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2014-11-01

    The fuel development program for research and test reactors calls for improved knowledge on the effect of microstructure on fuel performance in reactors. This paper summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 5.2 1021 fissions/cm3. While a large fraction of the fuel grains is decorated with large bubbles, there is no evidence showing interlinking of these bubbles at the specified fission density. The attachment of solid fission product precipitates to the bubbles is likely the result of fission product diffusion into these bubbles. The process of fission gas bubble superlattice collapse appears through bubble coalescence. The results are compared with the previous TEM work on the dispersion fuels irradiated to lower fission density from the same fuel plate.

  3. Competition of fusion and quasi-fission in the reactions leading to production of the superheavy elements

    E-print Network

    M. Veselsky

    2003-02-11

    The mechanism of fusion hindrance, an effect observed in the reactions of cold, warm and hot fusion leading to production of the superheavy elements, is investigated. A systematics of transfermium production cross sections is used to determine fusion probabilities. Mechanism of fusion hindrance is described as a competition of fusion and quasi-fission. Available evaporation residue cross sections in the superheavy region are reproduced satisfactorily. Analysis of the measured capture cross sections is performed and a sudden disappearance of the capture cross sections is observed at low fusion probabilities. A dependence of the fusion hindrance on the asymmetry of the projectile-target system is investigated using the available data. The most promising pathways for further experiments are suggested.

  4. Recoil-alpha-fission and recoil-alpha-alpha-fission events observed in the reaction Ca-48 + Am-243

    E-print Network

    U. Forsberg; D. Rudolph; L. -L. Andersson; A. Di Nitto; Ch. E. Dllmann; J. M. Gates; P. Golubev; K. E. Gregorich; C. J. Gross; R. -D. Herzberg; F. P. Hessberger; J. Khuyagbaatar; J. V. Kratz; K. Rykaczewski; L. G. Sarmiento; M. Schdel; A. Yakushev; S. berg; D. Ackermann; M. Block; H. Brand; B. G. Carlsson; D. Cox; X. Derkx; J. Dobaczewski; K. Eberhardt; J. Even; C. Fahlander; J. Gerl; E. Jger; B. Kindler; J. Krier; I. Kojouharov; N. Kurz; B. Lommel; A. Mistry; C. Mokry; W. Nazarewicz; H. Nitsche; J. P. Omtvedt; P. Papadakis; I. Ragnarsson; J. Runke; H. Schaffner; B. Schausten; Y. Shi; P. Thrle-Pospiech; T. Torres; T. Traut; N. Trautmann; A. Trler; A. Ward; D. E. Ward; N. Wiehl

    2015-02-10

    Products of the fusion-evaporation reaction Ca-48 + Am-243 were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum f\\"ur Schwerionenforschung. Amongst the detected thirty correlated alpha-decay chains associated with the production of element Z=115, two recoil-alpha-fission and five recoil-alpha-alpha-fission events were observed. The latter are similar to four such events reported from experiments performed at the Dubna gas-filled separator. Contrary to their interpretation, we propose an alternative view, namely to assign eight of these eleven decay chains of recoil-alpha(-alpha)-fission type to start from the 3n-evaporation channel 115-288. The other three decay chains remain viable candidates for the 2n-evaporation channel 115-289.

  5. Recoil-alpha-fission and recoil-alpha-alpha-fission events observed in the reaction Ca-48 + Am-243

    E-print Network

    Forsberg, U; Andersson, L -L; Di Nitto, A; Dllmann, Ch E; Gates, J M; Golubev, P; Gregorich, K E; Gross, C J; Herzberg, R -D; Hessberger, F P; Khuyagbaatar, J; Kratz, J V; Rykaczewski, K; Sarmiento, L G; Schdel, M; Yakushev, A; berg, S; Ackermann, D; Block, M; Brand, H; Carlsson, B G; Cox, D; Derkx, X; Dobaczewski, J; Eberhardt, K; Even, J; Fahlander, C; Gerl, J; Jger, E; Kindler, B; Krier, J; Kojouharov, I; Kurz, N; Lommel, B; Mistry, A; Mokry, C; Nazarewicz, W; Nitsche, H; Omtvedt, J P; Papadakis, P; Ragnarsson, I; Runke, J; Schaffner, H; Schausten, B; Shi, Y; Thrle-Pospiech, P; Torres, T; Traut, T; Trautmann, N; Trler, A; Ward, A; Ward, D E; Wiehl, N

    2015-01-01

    Products of the fusion-evaporation reaction Ca-48 + Am-243 were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum f\\"ur Schwerionenforschung. Amongst the detected thirty correlated alpha-decay chains associated with the production of element Z=115, two recoil-alpha-fission and five recoil-alpha-alpha-fission events were observed. The latter are similar to four such events reported from experiments performed at the Dubna gas-filled separator. Contrary to their interpretation, we propose an alternative view, namely to assign eight of these eleven decay chains of recoil-alpha(-alpha)-fission type to start from the 3n-evaporation channel 115-288. The other three decay chains remain viable candidates for the 2n-evaporation channel 115-289.

  6. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    SciTech Connect

    Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Dewayatna, Winter, E-mail: winter@batan.go.id [Center for Nuclear Fuel Technology, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Yos, E-mail: soj@batan.go.id [Center for Nuclear Equipment and Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

    2014-09-30

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g ( 10{sup 6} cm{sup ?1}) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g ( 10{sup 6} cm{sup ?1}) the limits were exceeded.

  7. Spectroscopy of few-particle nuclei around magic {sup 132}Sn from fission product {gamma}-ray studies.

    SciTech Connect

    Zhang, C. T.

    1998-07-29

    We are studying the yrast structure of very neutron-rich nuclei around doubly magic {sup 132}Sn by analyzing fission product {gamma}-ray data from a {sup 248}Cm source at Eurogam II. Yrast cascades in several few-valence-particle nuclei have been identified through {gamma}{gamma} cross coincidences with their complementary fission partners. Results for two-valence-particle nuclei {sup 132}Sb, {sup 134}Te, {sup 134}Sb and {sup 134}Sn provide empirical nucleon-nucleon interactions which, combined with single-particle energies already known in the one-particle nuclei, are essential for shell-model analysis in this region. Findings for the N = 82 nuclei {sup 134}Te and {sup 135}I have now been extended to the four-proton nucleus {sup 136}Xe. Results for the two-neutron nucleus {sup 134}Sn and the N = 83 isotones {sup 134}Sb, {sup 135}Te and {sup 135}I open up the spectroscopy of nuclei in the northeast quadrant above {sup 132}Sn.

  8. Mitochondrial fusion but not fission regulates larval growth and synaptic development through steroid hormone production

    PubMed Central

    Sandoval, Hector; Yao, Chi-Kuang; Chen, Kuchuan; Jaiswal, Manish; Donti, Taraka; Lin, Yong Qi; Bayat, Vafa; Xiong, Bo; Zhang, Ke; David, Gabriela; Charng, Wu-Lin; Yamamoto, Shinya; Duraine, Lita; Graham, Brett H; Bellen, Hugo J

    2014-01-01

    Mitochondrial fusion and fission affect the distribution and quality control of mitochondria. We show that Marf (Mitochondrial associated regulatory factor), is required for mitochondrial fusion and transport in long axons. Moreover, loss of Marf leads to a severe depletion of mitochondria in neuromuscular junctions (NMJs). Marf mutants also fail to maintain proper synaptic transmission at NMJs upon repetitive stimulation, similar to Drp1 fission mutants. However, unlike Drp1, loss of Marf leads to NMJ morphology defects and extended larval lifespan. Marf is required to form contacts between the endoplasmic reticulum and/or lipid droplets (LDs) and for proper storage of cholesterol and ecdysone synthesis in ring glands. Interestingly, human Mitofusin-2 rescues the loss of LD but both Mitofusin-1 and Mitofusin-2 are required for steroid-hormone synthesis. Our data show that Marf and Mitofusins share an evolutionarily conserved role in mitochondrial transport, cholesterol ester storage and steroid-hormone synthesis. DOI: http://dx.doi.org/10.7554/eLife.03558.001 PMID:25313867

  9. Ground movements associated with gas hydrate production

    SciTech Connect

    Siriwardane, H.J.

    1992-10-01

    The mechanics of ground movements during hydrate production can be more closely simulated by considering similarities with ground movements associated with subsidence in permafrost regions than with gob compaction in a longwall mine. The purpose of this research work is to investigate the potential strata movements associated with hydrate production by considering similarities with ground movements in permafrost regions. The work primarily involves numerical modeling of subsidence caused by hydrate dissociation. The investigation is based on the theories of continuum mechanics , thermomechanical behavior of frozen geo-materials, and principles of rock mechanics and geomechanics. It is expected that some phases of the investigation will involve the use of finite element method, which is a powerful computer-based method which has been widely used in many areas of science and engineering. Parametric studies will be performed to predict expected strata movements and surface subsidence for different reservoir conditions and properties of geological materials. The results from this investigation will be useful in predicting the magnitude of the subsidence problem associated with gas hydrate production. The analogy of subsidence in permafrost regions may provide lower bounds for subsidence expected in hydrate reservoirs. Furthermore, it is anticipated that the results will provide insight into planning of hydrate recovery operations.

  10. Gas Production Strategy of Underground Coal Gasification Based on Multiple Gas Sources

    PubMed Central

    Tianhong, Duan; Zuotang, Wang; Limin, Zhou; Dongdong, Li

    2014-01-01

    To lower stability requirement of gas production in UCG (underground coal gasification), create better space and opportunities of development for UCG, an emerging sunrise industry, in its initial stage, and reduce the emission of blast furnace gas, converter gas, and coke oven gas, this paper, for the first time, puts forward a new mode of utilization of multiple gas sources mainly including ground gasifier gas, UCG gas, blast furnace gas, converter gas, and coke oven gas and the new mode was demonstrated by field tests. According to the field tests, the existing power generation technology can fully adapt to situation of high hydrogen, low calorific value, and gas output fluctuation in the gas production in UCG in multiple-gas-sources power generation; there are large fluctuations and air can serve as a gasifying agent; the gas production of UCG in the mode of both power and methanol based on multiple gas sources has a strict requirement for stability. It was demonstrated by the field tests that the fluctuations in gas production in UCG can be well monitored through a quality control chart method. PMID:25114953

  11. Gas production strategy of underground coal gasification based on multiple gas sources.

    PubMed

    Tianhong, Duan; Zuotang, Wang; Limin, Zhou; Dongdong, Li

    2014-01-01

    To lower stability requirement of gas production in UCG (underground coal gasification), create better space and opportunities of development for UCG, an emerging sunrise industry, in its initial stage, and reduce the emission of blast furnace gas, converter gas, and coke oven gas, this paper, for the first time, puts forward a new mode of utilization of multiple gas sources mainly including ground gasifier gas, UCG gas, blast furnace gas, converter gas, and coke oven gas and the new mode was demonstrated by field tests. According to the field tests, the existing power generation technology can fully adapt to situation of high hydrogen, low calorific value, and gas output fluctuation in the gas production in UCG in multiple-gas-sources power generation; there are large fluctuations and air can serve as a gasifying agent; the gas production of UCG in the mode of both power and methanol based on multiple gas sources has a strict requirement for stability. It was demonstrated by the field tests that the fluctuations in gas production in UCG can be well monitored through a quality control chart method. PMID:25114953

  12. Mass-yield distributions of fission products from photofission of 232Th induced by 45- and 80-MeV bremsstrahlung

    NASA Astrophysics Data System (ADS)

    Naik, H.; Goswami, A.; Kim, G. N.; Lee, M. W.; Kim, K. S.; Suryanarayana, S. V.; Kim, E. A.; Shin, S. G.; Cho, M.-H.

    2012-11-01

    The mass-yield distributions of various fission products in the 45- and 80-MeV bremsstrahlung-induced fission of 232Th have been determined by using a recoil catcher and an offline ?-ray spectrometric technique in the electron linac at the Pohang Accelerator Laboratory, Korea. The mass-yield distributions were obtained from the fission-product yield data using charge-distribution corrections. The peak-to-valley (P/V) ratio, the average value of light mass () and heavy mass (), and the average number of neutrons () in the bremsstrahlung-induced fission of 232Th at different excitation energies were obtained from the mass-yield data. From the present measurements and the existing data from the 232Th(?,f) reaction and those from the 232Th(n,f) reaction at various energies, the following observations were obtained: (i) The mass-yield distributions in the 232Th(?,f) reaction at various energies are triple humped, similar to those of the 232Th(n,f) reaction. (ii) The yields of fission products for A = 133-134, A = 138-139, and A = 143-144 and their complementary products in the 232Th(?,f) reaction are higher than those of other fission products due to the nuclear structure effect. (iii) The yields of symmetric fission products for A = 133-134 and their complementary products in the 232Th(?,f) reaction are lower than those in the 232Th(n,f) reaction, whereas those for A = 143-144 and their complementary products are reversed. (iv) The result of increasing of the symmetric product yield causes the decreasing of the peak-to-valley ratio with increasing the excitation energy. However, it is surprising to see that the increasing trends for the symmetric products yields and the decreasing trends for the P/V ratio in the 232Th(?,f) and 232Th(n,f) reactions are not similar but those in the 238U(?,f) and 238U(n,f) reactions are similar to each other. (v) The average values of , , and at different excitation energies in the 232Th(?,f) and 232Th(n,f) reactions are similar but those in the 238U(?,f) and 238U(n,f) reactions are different.

  13. Delayed fission product gamma-ray transmission through low enriched uranium dioxide fuel pin lattices in air

    NASA Astrophysics Data System (ADS)

    Trumbull, Timothy H.

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called "channeling" effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice. Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins in an air medium. The calculational results support the measurements and suggest that under certain conditions material homogenization of fuel pin lattices does not introduce significant error in the dose rate calculation.

  14. Production of CoQ10 in fission yeast by expression of genes responsible for CoQ10 biosynthesis.

    PubMed

    Moriyama, Daisuke; Hosono, Kouji; Fujii, Makoto; Washida, Motohisa; Nanba, Hirokazu; Kaino, Tomohiro; Kawamukai, Makoto

    2015-06-01

    Coenzyme Q10 (CoQ10) is essential for energy production and has become a popular supplement in recent years. In this study, CoQ10 productivity was improved in the fission yeast Schizosaccharomyces pombe. Ten CoQ biosynthetic genes were cloned and overexpressed in S. pombe. Strains expressing individual CoQ biosynthetic genes did not produce higher than a 10% increase in CoQ10 production. In addition, simultaneous expression of all ten coq genes did not result in yield improvements. Genes responsible for the biosynthesis of p-hydroxybenzoate and decaprenyl diphosphate, both of which are CoQ biosynthesis precursors, were also overexpressed. CoQ10 production was increased by overexpression of Eco_ubiC (encoding chorismate lyase), Eco_aroF(FBR) (encoding 3-deoxy-d-arabino-heptulosonate 7-phosphate synthase), or Sce_thmgr1 (encoding truncated HMG-CoA reductase). Furthermore, simultaneous expression of these precursor genes resulted in two fold increases in CoQ10 production. PMID:25647499

  15. Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA

    E-print Network

    J. Diaz Leon; D. A. Jaffe; J. Kaspar; A. Knecht; M. L. Miller; R. G. H. Robertson; A. G. Schubert

    2011-08-23

    We report results of air monitoring started due to the recent natural catastrophe on 11 March 2011 in Japan and the severe ensuing damage to the Fukushima Dai-ichi nuclear reactor complex. On 17-18 March 2011, we registered the first arrival of the airborne fission products 131-I, 132-I, 132-Te, 134-Cs, and 137-Cs in Seattle, WA, USA, by identifying their characteristic gamma rays using a germanium detector. We measured the evolution of the activities over a period of 23 days at the end of which the activities had mostly fallen below our detection limit. The highest detected activity amounted to 4.4 +/- 1.3 mBq/m^3 of 131-I on 19-20 March.

  16. Fission-product yield data from the US/UK joint experiment in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Dickens, J.K.; Raman, S.

    1986-04-01

    The United States and the United Kingdom have been engaged in a joint research program in which samples of fissile and fertile actinides have been incorporated in fuel pins and irradiated in the Dounreay Prototype Fast Reactor in Scotland. The purpose of this portion of the program is to study both the materials behavior and the nuclear physics results - primarily measurements of the fission-product yields in the irradiated samples and secondarily information on the amounts of heavy elements in the samples. In the measurements high-resolution detectors were used to observe and (quantitatively measure) the gamma rays and x rays corresponding to the decay of several long-lived radioisotopes. Two series of measurements were made, one nine months following the end of the irradiation period and another approximately six months later.

  17. Collection of fission and activation product elements from fresh and ocean waters: a comparison of traditional and novel sorbents

    SciTech Connect

    Johnson, Bryce E.; Santschi, Peter H.; Addleman, Raymond S.; Douglas, Matthew; Davidson, Joseph D.; Fryxell, Glen E.; Schwantes, Jon M.

    2010-04-01

    Monitoring natural waters for the inadvertent release of radioactive fission products produced as a result of nuclear power generation downstream from these facilities is essential for maintaining water quality. To this end, we evaluated sorbents for simultaneous in-situ large volume extraction of radionuclides with both soft (e.g., Ag) and hard metal (e.g., Co, Zr, Nb, Ba, and Cs) or anionic (e.g., Ru, Te, Sb) character. In this study, we evaluated a number of conventional and novel nanoporous sorbents in both fresh and salt waters. In most cases, the nanoporous sorbents demonstrated enhanced retention of analytes. Salinity had significant effects upon sorbent performance and was most significant for hard cations, specifically Cs and Ba. The presence of natural organic matter had little effect on the ability of chemisorbents to extract target elements.

  18. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    PubMed

    Ko?l, Michal; vadlenkov, Marie; Mil?k, Jn; Rypar, Vojt?ch; Koleka, Michal

    2014-07-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well. PMID:24566373

  19. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (inventors)

    1982-01-01

    A thermal reactor apparatus and method of pyrolyticaly decomposing silane gas into liquid silicon product and hydrogen by-product gas is disclosed. The thermal reactor has a reaction chamber which is heated well above the decomposition temperature of silane. An injector probe introduces the silane gas tangentially into the reaction chamber to form a first, outer, forwardly moving vortex containing the liquid silicon product and a second, inner, rewardly moving vortex containing the by-product hydrogen gas. The liquid silicon in the first outer vortex deposits onto the interior walls of the reaction chamber to form an equilibrium skull layer which flows to the forward or bottom end of the reaction chamber where it is removed. The by-product hydrogen gas in the second inner vortex is removed from the top or rear of the reaction chamber by a vortex finder. The injector probe which introduces the silane gas into the reaction chamber is continually cooled by a cooling jacket.

  20. 30 CFR 202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ...determine the royalty due on gas production? 202.550 Section 202...REVENUE MANAGEMENT ROYALTIES Gas Production From Indian Leases 202.550...determine the royalty due on gas production? If you produce gas...

  1. Direct quantitative determination, using flameless atomic absorption spectroscopy, of metallic impurities and rare earths in nuclear solutions containing uranium, thorium, and fission products

    Microsoft Academic Search

    M. Gerardi; G. A. Pelliccia

    2009-01-01

    A method is described for direct quantitative determination of metallic and rare earths in nuclear solutions containing U, Th, and fission products from the ITREC reprocessing plant of the CNEN, CRN-Trisaia. Various tables summarize instrumental operating conditions, preparations of standards, detection limits and sensitivities, average relative standard deviation, analytical limits for the 22 investigated elements, maximum total tolerance of the

  2. Review of areas that may require simultaneous coupled solution of the thermal hydraulic and fission product\\/aerosol behavior equations for source term determination. [PWR; BWR

    Microsoft Academic Search

    Kress

    1984-01-01

    In the determination of the behavior of nuclear aerosols in the reactor coolant system and in the containment for the development of severe accident source terms, present practice generally is to first perform thermal hydraulic calculations for specific plant types and sequences and then to utilize the results as input for separate fission product\\/aerosol dynamic transport calculations. It is recognized

  3. A study of fission product migration and selective leaching by means of a power-bump test

    SciTech Connect

    Forsyth, R.S. [Caledon-Consult, Nykoeping (Sweden); Eklund, U.B. [Studsvik Nuclear AB, Nykoeping (Sweden); Werme, L.O. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    1994-12-31

    As part of a programme to identify and quantify the mechanisms controlling the aqueous corrosion of spent fuel, a short fuel rod was refabricated from a full-length rod from the Swedish Ringhals 1 BWR (burnup 44-48 MWd/kgU). After a conditioning period at 25 kW/m, the rod was power-bumped in Studsvik Nuclear`s R2 test reactor up to a maximum linear heat rating of 43 kW/m, with a hold time at power of 3 hours, imposing steeper radial temperature gradients on the fuel pellets. A few hours after termination of irradiation, the rod was shipped to Studsvik Nuclear`s Hot Cell Laboratory, where the pellet inventories of gamma-emitting fission products were determined non-destructively. Two 20 mm long fuel/clad sections were cut from the rod at positions corresponding to power-bump LHR values of 43 and 33 kW/m. Two similar sections were cut from the original rod as references, together with a shorter section for destructive inventory analysis. Fractional release data for both radioactive and stable fission products has been obtained. Since the main aim of the test was the study of the early stages of corrosion, expected to be dominated by migrational effects in the fuel during irradiation, particular attention has been paid to Cs, I, Rb, Sr, Ba, Mo and Tc. Although the migration and selective leaching of Cs and I are well-established, it is shown that there is also appreciable leaching of Rb nuclides, whereas the results show no significant Sr migration. This represents support for the hypothesis that the apparent enhanced leach rate of Sr-90 observed experimentally can be due to migration of Rb-90 precursors.

  4. FISSION GAS SAMPLER

    Microsoft Academic Search

    Chismar

    1963-01-01

    An apparatus was developed for measuring and sampling free gases in ; uranium oxide fuel elements. The apparatus was designed to release the gases ; contained in the element and collect a fraction of the gases for analysis. ; (auth);

  5. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion exchange during the salt/zeolite contacting process Compare the adsorption models to experimentally obtained, ER salt results Evaluate results obtained from the oxygen precipitation and salt/zeolite ion exchange studies to determine the best processes for selective fission-product removal from electrorefiner salt.

  6. GASCAP: Wellhead Gas Productive Capacity Model documentation, June 1993

    SciTech Connect

    Not Available

    1993-07-01

    The Wellhead Gas Productive Capacity Model (GASCAP) has been developed by EIA to provide a historical analysis of the monthly productive capacity of natural gas at the wellhead and a projection of monthly capacity for 2 years into the future. The impact of drilling, oil and gas price assumptions, and demand on gas productive capacity are examined. Both gas-well gas and oil-well gas are included. Oil-well gas productive capacity is estimated separately and then combined with the gas-well gas productive capacity. This documentation report provides a general overview of the GASCAP Model, describes the underlying data base, provides technical descriptions of the component models, diagrams the system and subsystem flow, describes the equations, and provides definitions and sources of all variables used in the system. This documentation report is provided to enable users of EIA projections generated by GASCAP to understand the underlying procedures used and to replicate the models and solutions. This report should be of particular interest to those in the Congress, Federal and State agencies, industry, and the academic community, who are concerned with the future availability of natural gas.

  7. Predicting the production of neutron-rich heavy nuclei in multinucleon transfer reactions using a semi-classical model including evaporation and fission competition, GRAZING-F

    NASA Astrophysics Data System (ADS)

    Yanez, R.; Loveland, W.

    2015-04-01

    Background: Multinucleon transfer reactions have recently attracted attention as a possible path to the synthesis of new neutron-rich heavy nuclei. Purpose: We study transfer reactions involving massive nuclei with the intention of understanding if the semi-classical model GRAZING coupled to an evaporation and fission competition model can satisfactorily reproduce experimental data on transfer reactions in which fission plays a role. Methods: We have taken the computer code grazing and have added fission competition to it (grazing-f) using our current understanding of ?n/?f , fission barriers, and level densities. Results: The code grazing-f seems to satisfactorily reproduce experimental data for +1 p ,+2 p , and +3 p transfers but has limitations in reproducing measurements of larger above-target and below-target transfers. Nonetheless, we use grazing-f to estimate production rates of neutron-rich N =126 nuclei, actinides, and transactinides. Conclusions: The grazing code, with appropriate modifications to account for fission decay as well as neutron emission by excited primary fragments, does not predict large cross sections for multinucleon transfer reactions leading to neutron-rich transactinide nuclei but predicts opportunities to produce new neutron-rich actinide isotopes.

  8. Measurement of fission gas release rates from a uranium dioxide fuel pin during irradiation in a fast reactor

    Microsoft Academic Search

    W. B. Bremner; A. B. G. Washington

    1973-01-01

    An experiment carried out in the Dounreay fast reactor to measure ; continuously the fission gases released from a long uranium dioxide fuel pin ; irradiated to a burn-up of 2% is described. The gases were swept by a helium ; purge from the pin to the reactor top where they were analyzed by bottle sampling ; and in line

  9. Monthly Crude Oil and Natural Gas Production Report

    EIA Publications

    2015-01-01

    Natural gas production (gross withdrawals) from data collected on Form EIA-914 (Monthly Crude Oil, Lease Condensate, and Natural Gas Production Report) for Federal Offshore Gulf of Mexico, Texas, Louisiana, New Mexico, Oklahoma, Texas, Wyoming, other states and lower 48 states. Alaska data are from the Alaska state government and included to obtain a U.S. total.

  10. Designing an upgrade of the Medley setup for light-ion production and fission cross-section measurements

    E-print Network

    Kaj Jansson; Cecilia Gustavsson; Ali Al-Adili; Anders Hjalmarsson; Erik Andersson-Sundn; Alexander V. Prokofiev; Diego Tarro; Stephan Pomp

    2015-06-23

    Measurements of neutron-induced fission cross sections and light-ion production are planned in the energy range 1-40 MeV at the upcoming Neutrons For Science (NFS) facility. In order to prepare our detector setup for the neutron beam with continuous energy spectrum, a simulation software was written using the Geant4 toolkit for both measurement situations. The neutron energy range around 20 MeV is troublesome when it comes to the cross sections used by Geant4 since data-driven cross sections are only available below 20 MeV but not above, where they are based on semi-empirical models. Several customisations were made to the standard classes in Geant4 in order to produce consistent results over the whole simulated energy range. Expected uncertainties are reported for both types of measurements. The simulations have shown that a simultaneous precision measurement of the three standard cross sections H(n,n), $^{235}$U(n,f) and $^{238}$U(n,f) relative to each other is feasible using a triple layered target. As high resolution timing detectors for fission fragments we plan to use Parallel Plate Avalanche Counters (PPACs). The simulation results have put some restrictions on the design of these detectors as well as on the target design. This study suggests a fissile target no thicker than 2 micrometers (1.7 mg/cm$^2$) and a PPAC foil thickness preferably less than 1 micrometer. We also comment on the usability of Geant4 for simulation studies of neutron reactions in this energy range.

  11. Production of biodiesel using expanded gas solvents

    SciTech Connect

    Ginosar, Daniel M [Idaho Falls, ID; Fox, Robert V [Idaho Falls, ID; Petkovic, Lucia M [Idaho Falls, ID

    2009-04-07

    A method of producing an alkyl ester. The method comprises providing an alcohol and a triglyceride or fatty acid. An expanding gas is dissolved into the alcohol to form a gas expanded solvent. The alcohol is reacted with the triglyceride or fatty acid in a single phase to produce the alkyl ester. The expanding gas may be a nonpolar expanding gas, such as carbon dioxide, methane, ethane, propane, butane, pentane, ethylene, propylene, butylene, pentene, isomers thereof, and mixtures thereof, which is dissolved into the alcohol. The gas expanded solvent may be maintained at a temperature below, at, or above a critical temperature of the expanding gas and at a pressure below, at, or above a critical pressure of the expanding gas.

  12. TOAFEW-V multigroup cross-section collapsing code and library of 154-group-processed ENDF\\/B\\/V fission-product and actinide cross sections

    Microsoft Academic Search

    W. B. Wilson; T. R. England; R. J. LaBauve; R. M. Boicourt

    1982-01-01

    The ENDF\\/B-V cross sections of 237 fission-product and actinide nuclides have been processed at infinite dilution into 154 neutron energy groups at temperatures of 300, 900, and 1200 K. Additional 154-group self-shielded actinide cross sections were calculated with Bondarenko background sigma values as small as 1 barn. The production and content of the multigroup data library is described. The TOAFEW-V

  13. Potentials of fissioning plasmas

    NASA Technical Reports Server (NTRS)

    Thom, K.

    1979-01-01

    Successful experiments with the nuclear pumping of lasers have demonstrated that in a gaseous medium the kinetic energy of fission fragments can be converted directly into nonequilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in near thermal equilibrium varying up to a state of extreme nonequilibrium. During 20 years of research under NASA support major elements of the fissioning plasma reactor were demonstrated in theory and experiment, culminating in a proof-of-principle reactor test conducted at the Los Alamos Scientific Laboratory. It is concluded that the construction of a gaseous fuel reactor power plant is within the reach of present technology.

  14. Ionizing radiation accelerates Drp1-dependent mitochondrial fission, which involves delayed mitochondrial reactive oxygen species production in normal human fibroblast-like cells

    SciTech Connect

    Kobashigawa, Shinko, E-mail: kobashin@nagasaki-u.ac.jp [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)] [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan); Suzuki, Keiji; Yamashita, Shunichi [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)] [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)

    2011-11-04

    Highlights: Black-Right-Pointing-Pointer We report first time that ionizing radiation induces mitochondrial dynamic changes. Black-Right-Pointing-Pointer Radiation-induced mitochondrial fission was caused by Drp1 localization. Black-Right-Pointing-Pointer We found that radiation causes delayed ROS from mitochondria. Black-Right-Pointing-Pointer Down regulation of Drp1 rescued mitochondrial dysfunction after radiation exposure. -- Abstract: Ionizing radiation is known to increase intracellular level of reactive oxygen species (ROS) through mitochondrial dysfunction. Although it has been as a basis of radiation-induced genetic instability, the mechanism involving mitochondrial dysfunction remains unclear. Here we studied the dynamics of mitochondrial structure in normal human fibroblast like cells exposed to ionizing radiation. Delayed mitochondrial O{sub 2}{sup {center_dot}-} production was peaked 3 days after irradiation, which was coupled with accelerated mitochondrial fission. We found that radiation exposure accumulated dynamin-related protein 1 (Drp1) to mitochondria. Knocking down of Drp1 expression prevented radiation induced acceleration of mitochondrial fission. Furthermore, knockdown of Drp1 significantly suppressed delayed production of mitochondrial O{sub 2}{sup {center_dot}-}. Since the loss of mitochondrial membrane potential, which was induced by radiation was prevented in cells knocking down of Drp1 expression, indicating that the excessive mitochondrial fission was involved in delayed mitochondrial dysfunction after irradiation.

  15. Integrated production of fuel gas and oxygenated organic compounds from synthesis gas

    DOEpatents

    Moore, Robert B. (Allentown, PA); Hegarty, William P. (State College, PA); Studer, David W. (Wescosville, PA); Tirados, Edward J. (Easton, PA)

    1995-01-01

    An oxygenated organic liquid product and a fuel gas are produced from a portion of synthesis gas comprising hydrogen, carbon monoxide, carbon dioxide, and sulfur-containing compounds in a integrated feed treatment and catalytic reaction system. To prevent catalyst poisoning, the sulfur-containing compounds in the reactor feed are absorbed in a liquid comprising the reactor product, and the resulting sulfur-containing liquid is regenerated by stripping with untreated synthesis gas from the reactor. Stripping offgas is combined with the remaining synthesis gas to provide a fuel gas product. A portion of the regenerated liquid is used as makeup to the absorber and the remainder is withdrawn as a liquid product. The method is particularly useful for integration with a combined cycle coal gasification system utilizing a gas turbine for electric power generation.

  16. Pol5p, a novel binding partner to Cdc10p in fission yeast involved in rRNA production.

    PubMed

    Nadeem, Farzana Khaliq; Blair, Derek; McInerny, Christopher J

    2006-10-01

    Cdc10p is a major component of the cell cycle transcription factor complex MBF that controls G1-S phase specific gene expression in the fission yeast Schizosaccharomyces pombe. Here, we describe the identification of a new binding partner to Cdc10p and Pol5p. Pol5p was discovered through a 2-hybrid screen, with the direct interaction confirmed by in vitro "pull-down" experiments with bacterially expressed proteins. Pol5p appears to have no role in cell cycle gene expression, but is instead required for rRNA production. Pol5p is an essential gene, expressed constitutively throughout both the mitotic and meiotic life cycles, and localises to the nucleus. Over-expressing Pol5p has no phenotype, but reducing levels of Pol5p inhibits rRNA production. Pol5p is shown to bind to rDNA promoter fragments. Potentially, we have identified a mechanism by which Cdc10p controls rDNA gene expression, therefore linking the cell cycle with cellular growth. PMID:16816948

  17. Improving oil and gas production with the Beam-Mounted Gas Compressor

    SciTech Connect

    Al-Khatib, A.M.

    1984-02-01

    This paper explores parameters involved in and advantages obtained by use of the Beam-Mounted Gas Compressor (BMGC), a single-acting gas compressor operated by the walking beam of a rod pumping unit. Its main function is to draw gas from the casing side of an oil well and to discharge the gas into the flow line. By doing so, the BMGC operation actually reduces the backpressure on the formation face, thus allowing additional oil to enter the wellbore for production.

  18. Methane hydrate gas production: evaluating and exploiting the solid gas resource

    SciTech Connect

    McGuire, P.L.

    1981-01-01

    Methane hydrate gas could be a tremendous energy resource if methods can be devised to produce this gas economically. This paper examines two methods of producing gas from hydrate deposits by the injection of hot water or steam, and also examines the feasibility of hydraulic fracturing and pressure reduction as a hydrate gas production technique. A hydraulic fracturing technique suitable for hydrate reservoirs and a system for coring hydrate reservoirs are also described.

  19. Natural gas hydrates - issues for gas production and geomechanical stability

    E-print Network

    Grover, Tarun

    2008-10-10

    bearing sediments in offshore environments, I divided these data into different sections. The data included water depths, pore water salinity, gas compositions, geothermal gradients, and sedimentary properties such as sediment type, sediment mineralogy... .................................................................. 9 2.2 Hydrate patterns in sediments .................................................................... 24 3.1 Water depths and penetration for the Blake Ridge..................................... 31 3.2 Geothermal gradients measured...

  20. Local and Medium Range Order Around Fission Products in Inactive Waste Glasses: Implication for Glass Structure and Stability

    NASA Astrophysics Data System (ADS)

    Galoisy, L.; Calas, G.; Ghaleb, D.; Morin, G.

    2002-12-01

    Borosilicate glasses are used to store high level nuclear waste in France (R7T7 glass). The structure of the glass around elements such as fission products controls important parameters as the homogeneity of the glass and/or the melted glass rheology. Data on the local and medium range order structure of these glasses could help improving the resistance toward leaching and/or irradiation, in relation with surface or geological storage of these vitrified wastes. Due to the complex composition of these glasses (up to 30 oxides), chemically selective methods are required to understand the environment of elements. X-ray Absorption Spectroscopy (XAS) is, from this point of view, a powerful tool as it provides a direct access to the investigation of the structure around specific cations in this multicomponent amorphous material, to specify their role in the glass durability. We will present different XAS studies (synchrotrons in LURE and ESRF, France) on the inactive amorphous analog for the R7T7 glass (the SON 68 glass). This report will illustrate the potentialities of this approach through the determination of the environment around fission products such as Zr, Zn and Mo. XAS shows the peculiarity of the sites occupied by these glass components of technological interest. Coordination numbers are shown to be systematically smaller than in crystalline compounds with close composition. Below the definition of the sites occupied by the chemical elements, XAS allows to detect some degree of medium range order which gives insight on the bonding of the site to the poymeric borosilicate network and allow to link precisely experimental data to theoretical calculations. Eventually, XAS is used to study the interaction between noble metals (Pd and Ru) and the glassy matrix. These elements are at the origin of small precipitates that induce changes in the melt vicosity. They occur as a result of the non-insertion of these elements in the glassy matrix. To accurate and precise structural interpretations, a direct comparison with MD calculations on simplified nuclear glass comprising 5 oxides, is performed.

  1. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling methods used in this study.

  2. Chlorination kinetics for actinoid and fission-product phosphates in chloride melts. 3. Cerium phosphate

    SciTech Connect

    Kryukova, A.I.; Burnaeva, A.A.; Skiba, O.V.; Korshunov, I.A.

    1987-09-01

    Kinetic measurements have been made on the reactions of cerium phosphate with carbon tetrachloride vapor in the molten chlorides of sodium, potassium, rubidium, and cesium. There is a discussion of the effects of temperature, gas flow, liquid composition, and grain size of the phosphate solid on the chlorination rate. Effective rate constants and activation energies have been calculated. A model has been constructed for the chlorination by means of a complete factor experiment.

  3. Analysis of primary damage in silicon carbide under fusion and fission neutron spectra

    NASA Astrophysics Data System (ADS)

    Guo, Daxi; Zang, Hang; Zhang, Peng; Xi, Jianqi; Li, Tao; Ma, Li; He, Chaohui

    2014-12-01

    Irradiation parameters on primary damage states of SiC are evaluated and compared for the first wall of ITER under deuterium-deuterium (DD) and deuterium-tritium (DT) operation, the high temperature gas-cooled reactor (HTGR) and high flux isotope reactor (HFIR). With the same neutron fluence, the studied fusion spectra produce more damage and much higher gas production than the fission spectra. Due to comparable gas production and similar weighted primary recoil spectra, HFIR is considered suitable to simulate the neutron irradiation in an HTGR. In contrast to the significant differences between the weighted primary recoil spectra of the fission and the fusion spectra, the weighted secondary recoil spectra of HFIR and HTGR match those of DD and DT, indicating that displacement cascades by the fission and the fusion irradiation are similar when the damage distribution among damaged regions by secondary recoils is taken into account.

  4. Gas-phase terpene oxidation products: a review

    Microsoft Academic Search

    A. Calogirou; B. R. Larsen; D. Kotzias

    1999-01-01

    Terpenes are emitted in large quantities from vegetation into the troposphere, where they react readily with ozone, OH and NO3 radicals leading to a number of oxidation products. The current knowledge about gas-phase terpene oxidation products is reviewed. Their formation and decomposition pathways, their products and their relevance for the troposphere, and their chemical analysis are discussed. Data on oxidation

  5. Gas hearth products market fact base. Topical report, January 1996

    SciTech Connect

    NONE

    1996-02-01

    The Gas Hearth Products Market Fact Base is an analysis of the U.S. gas log and fireplace markets. The study was undertaken to: determine current usage of and attitudes about fireplaces; identify barriers to acceptance of gas logs and fireplaces; determine the influence of service providers, and; identify important trends that can affect the markets for gas hearth products. The market fact base is based on four studies: a market analysis synthesizing primary and secondary research reports; in-depth interviews with market influencers from across the country (architects, contractors, interior designers, fireplace retailers and installers) and industry experts from gas utilities and trade associations; focus group meetings with consumers who own or intend to buy fireplaces, gas fireplace industry professionals, and editors of fireplace-related trade magazines, and; quantitative interviews with consumers in six U.S. cities.

  6. Production Trends of Shale Gas Wells

    E-print Network

    Khan, Waqar A.

    2010-01-14

    different flow regions in shale gas wells that include linear and bilinear flow. Important field parameters can be calculated from those observations to help improve future performance. The detailed plots of several wells in this study show some good numbers...

  7. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...2013-07-01 false How do I measure natural gas production on my eligible lease... 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  8. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...2014-07-01 false How do I measure natural gas production on my eligible lease... 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  9. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...I determine the royalty due on gas production? 1202.550 Section...DEPARTMENT OF THE INTERIOR Natural Resources Revenue ROYALTIES Gas Production From Indian Leases...I determine the royalty due on gas production? If you...

  10. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...2011-07-01 false How do I measure natural gas production on my eligible lease... 260.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  11. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...2012-07-01 false How do I measure natural gas production on my eligible lease... 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  12. 76 FR 67201 - Information Collection Activities: Oil and Gas Production Safety Systems; Submitted for Office of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-31

    ...Activities: Oil and Gas Production Safety Systems; Submitted for Office of...Subpart H, ``Oil and Gas Production Safety Systems.'' This notice also provides...Subpart H, Oil and Gas Production Safety Systems. Abstract: The...

  13. Process for production desulfurized of synthesis gas

    Microsoft Academic Search

    James K. Wolfenbarger; Mitri S. Najjar

    1993-01-01

    A process for the partial oxidation of a sulfur- and silicate-containing carbonaceous fuel to produce a synthesis gas with reduced sulfur content which comprises partially oxidizing said fuel at a temperature in the range of 1900.degree.-2600.degree. F. in the presence of a temperature moderator, an oxygen-containing gas and a sulfur capture additive which comprises a calcium-containing compound portion, a sodium-containing

  14. A kinetic model for fission-product release and fuel oxidation behaviour for Zircaloy-clad fuel elements under reactor accident conditions

    NASA Astrophysics Data System (ADS)

    Lewis, B. J.; Cox, D. S.; Iglesias, F. C.

    1993-12-01

    An analytical model has been developed to describe the release behaviour of fission product cesium from uranium dioxide fuel during severe reactor accident conditions. The present methodology is based on the results of out-of-pile annealing experiments with irradiated fuel (bare and Zircaloy-clad fuel specimens), subjected to a steam atmosphere at high temperature (1200 to 1700C). In the present framework, the fuel oxidation kinetics is detailed by a surface exchange reaction at the fuel/steam interface. The dependence of the fission product release kinetics on the state of fuel oxidation is treated. This analysis also considers the inhibiting influence of hydrogen, produced as a result of the Zircaloy-steam reaction, on the oxygen potential of the steam environment.

  15. Wet deposition of fission-product isotopes to North America from the Fukushima Dai-ichi incident, March 2011

    USGS Publications Warehouse

    Wetherbee, Gregory A.; Gay, David A.; Debey, Timothy M.; Lehmann, Christopher M.B.; Nilles, Mark A.

    2012-01-01

    Using the infrastructure of the National Atmospheric Deposition Program (NADP), numerous measurements of radionuclide wet deposition over North America were made for 167 NADP sites before and after the Fukushima Dai-ichi Nuclear Power Station incident of March 12, 2011. For the period from March 8 through April 5, 2011, wet-only precipitation samples were collected by NADP and analyzed for fission-product isotopes within whole-water and filterable solid samples by the United States Geological Survey using gamma spectrometry. Variable amounts of 131I, 134Cs, or 137Cs were measured at approximately 21% of sampled NADP sites distributed widely across the contiguous United States and Alaska. Calculated 1- to 2-week individual radionuclide deposition fluxes ranged from 0.47 to 5100 Becquerels per square meter during the sampling period. Wet deposition activity was small compared to measured activity already present in U.S. soil. NADP networks responded to this complex disaster, and provided scientifically valid measurements that are comparable and complementary to other networks in North America and Europe.

  16. Ground-state ?--branching intensities of several fission-product isotopes measured using a total absorption ?-ray spectrometer

    NASA Astrophysics Data System (ADS)

    Greenwood, R. C.; Putnam, M. H.; Watts, K. D.

    1996-02-01

    The final set of results of "ground-state" ?--branching intensities obtained in a program of systematic study of those regions of the fission-product nuclides accessible to investigation using the 252Cf-based INEL ISOL facility are presented. A total absorption ?-ray spectrometer, operating in a 4??-? coincidence mode, was used to obtain these "ground-state" ?--branching intensities; where here the "ground-state" is defined to include all states below a selected ?-ray discriminator level. Results obtained for 89Rb, 90gRb, 91Rb, 93Rb, 93Sr, 94Sr, 94Y, 95Sr, 95Y, 140Cs, 142La, 143Ba, 143La, 144Ba, 144La, 145Ba, 145La, 146Ce, 146Pr, 147Ce, 147Pr, 148Ce, 148Pr, (2.27 min), 149Pr, 149Nd, 151Pr, 151Nd, 152Pm (4.1 min), 153Nd, 155Nd, 157Pm, 157Sm, 158Sm and 158Eu are presented and compared with existing published data.

  17. Measurement of ? --decay intensity distributions of several fission-product isotopes using a total absorption ?-ray spectrometer

    NASA Astrophysics Data System (ADS)

    Greenwood, R. C.; Helmer, R. G.; Putnam, M. H.; Watts, K. D.

    1997-02-01

    A total absorption ?-ray spectrometer coupled to the 252Cf-based INEL ISOL facility has been used in a program of systematic study of the distributions of ? --decay intensities of fission-product radionuclides. Cascade-summed ?-ray spectra measured with the system have been compared with the spectrum simulated from the corresponding decay schemes, as a test of the completeness and correctness of these schemes. New ? --decay intensity distributions have been deduced for the decay of these radionuclides. Radionuclides which have been studied in this manner include 89Rb, 90gRb, 90mRb, 91Rb, 93Rb, 93Sr, 94Sr, 94Y, 95Sr, 95Y, 138gCs, 138mCs, 139Cs, 140Cs, 141Cs, 141Ba, 142Ba, 142La, 143Ba, 143La, 144Ba, 144La, 145Ba, 145La, 145Ce, 146Ce, 146Pr, 147Ce, 147Pr, 148Ce, 148Pr (2.0 min), 148Pr (2.27 min), 149Pr, 149Nd, 151Pr, 151Nd, 152Nd - 152Pm (4.1 min.), 153Nd, 153Pm, 154Nd, 154Pm (1.7 min), 155Nd, 155Pm, 156Pm, 157Pm, 157Sm, 158Sm, and 158Eu.

  18. Nonsoluble fission product residues, crud, and fine chips of zircaloy cladding in headend process of nuclear fuel reprocessing

    SciTech Connect

    Gonda, K.; Hayashi, K.; Oka, K.

    1984-04-01

    The amount and behavior of fine suspended particles and sediments in headend process vessels were investigated. Powdery fines of Zircaloy cladding, crud, and nonsoluble fission product (FP) residues were determined to be 5.3, 1.8, and 1.0 kg/ton of spent fuel reprocessed, respectively. The 1.0 kg/ton of nonsoluble FP residues and 1.8 kg/ton of crud were reasonable amounts when compared with those estimated from burnup and amount of spent fuels treated. The 5.3 kg/ton spent fuel reprocessing came from powdery fines of Zircaloy cladding that had been confirmed by chop of unirradiated Zircaloy clad tube. These residues were mostly suspended in a process solution. Particle size of sediments and suspended particles distributed mostly in <0.5-mm size. Most of the particles that arose in the dissolver scarcely settled down and passed through headend process vessels into the high-active liquid waste storage vessel, while some of the particles settled down in succession in process vessels. Uranium and plutonium dissolved well, so that they left little nonsoluble residue. The weight fraction ratio of nonsoluble plutonium to uranium was 0.05% in sediments, which was higher than the value of 0.02% in hulls. It was concluded that uranium continues to dissolve even after settling down into sediments of the dissolver.

  19. Sensitivity of in-vessel hydrogen generation and fission product release to parameter variations in a melt progression model

    SciTech Connect

    Lee, M.; Khatib-Rahbar, M.

    1987-01-01

    The purpose of this paper is to determine the impact of the core meltdown model parameters on in-vessel hydrogen generation and radiological release for an anticipated transient without scram scenario in a boiling water reactor. As part of a continuing effort to quantify the radiological source term from a severe accident in a light water reactor, the US Nuclear Regulatory Commission (NRC) sponsored the development of the Source Term Code Package (STCP). In order to better establish the validity and potential applications of source term predictions from these codes, the Quantification and Uncertainty Analysis of Source Term for Severe Accidents in Light Water Reactors (QUASAR) program was initiated at Brookhaven National Lab. under the sponsorship of the NRC to quantify the uncertainties associated with the STCP calculated source term. The core melt progression model in the STCP is included in the MARCH code. The results of the sensitivity studies summarized. The in-vessel hydrogen generation and fission product release predictions are strongly influenced by the assumptions related to the core melt progression. Large uncertainties exist over a wide range of parametrics as shown in this paper.

  20. Fission products behavior in molten fluoride salts: Speciation of La3+ and Cs+ in melts containing oxide ions

    NASA Astrophysics Data System (ADS)

    Rollet, Anne-Laure; Veron, Emmanuel; Bessada, Catherine

    2012-10-01

    In this paper we address the effects of fission products on the speciation in molten fluoride salts. Numerous systems with cross-connections have been investigated in order to better identify the influence of CsF in a fluoride melt containing rare earth and oxides : LaF3-AF (A = Li, Na, K, Rb and Cs), LaF3-LiF-CsF, LaF3-LiF-CaF2, LaF3-LiF-CaO, LaF3-LiF-CaO-CsF. In this goal, we performed high temperature NMR experiments and followed in situ the evolution of 19F, 23Na, 85Rb, 133Cs and 139La NMR chemical shifts. In LaF3-AF-CsF and LaF3-AF-CaF2 systems, the coordination number of lanthanum cation ranges from 6 to 8 depending on the LaF3 concentration and on the polarizability of the other cations. The addition of oxide (CaO) in the latter mixtures leads to the formation of lanthanum oxyfluoride species that precipitate in LaOF when CaO concentration is increased. The addition of CsF to LaF3-LiF-CaO yields to a displacement of the dissolved versus precipitated LaOF proportion.

  1. Wet deposition of fission-product isotopes to North America from the Fukushima Dai-ichi incident, March 2011.

    PubMed

    Wetherbee, Gregory A; Gay, David A; Debey, Timothy M; Lehmann, Christopher M B; Nilles, Mark A

    2012-03-01

    Using the infrastructure of the National Atmospheric Deposition Program (NADP), numerous measurements of radionuclide wet deposition over North America were made for 167 NADP sites before and after the Fukushima Dai-ichi Nuclear Power Station incident of March 12, 2011. For the period from March 8 through April 5, 2011, wet-only precipitation samples were collected by NADP and analyzed for fission-product isotopes within whole-water and filterable solid samples by the United States Geological Survey using gamma spectrometry. Variable amounts of (131)I, (134)Cs, or (137)Cs were measured at approximately 21% of sampled NADP sites distributed widely across the contiguous United States and Alaska. Calculated 1- to 2-week individual radionuclide deposition fluxes ranged from 0.47 to 5100 Becquerels per square meter during the sampling period. Wet deposition activity was small compared to measured activity already present in U.S. soil. NADP networks responded to this complex disaster, and provided scientifically valid measurements that are comparable and complementary to other networks in North America and Europe. PMID:22356354

  2. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    SciTech Connect

    Brger, Stefan [New Brunswick Laboratory, Argonne, IL; Riciputi, Lee R [Los Alamos National Laboratory (LANL); Bostick, Debra A [ORNL; Turgeon, Steven [University of Alberta, Edmondton, Canada; McBay, Eddie H [ORNL; Lavelle, Mark [ORNL

    2009-01-01

    A ThermoFisher 'Triton' multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotoperatioanalysis of actinides (uranium, plutonium, and americium), fission products and geolocators (strontium, cesium, and neodymium). Total efficiencies (atoms loaded to ions detected) of up to 0.5-2% for U, Pu, and Am, and 1-30% for Sr, Cs, and Nd can be reported employing resin bead load techniques onto flat ribbon Re filaments or resin beads loaded into a millimeter-sized cavity drilled into a Re rod. This results in detection limits of <0.1 fg (10{sup 4} atoms to 10{sup 5} atoms) for {sup 239-242+244}Pu, {sup 233+236}U, {sup 241-243}Am, {sup 89,90}Sr, and {sup 134,135,137}Cs, and {le} 1 pg for natural Nd isotopes (limited by the chemical processing blank) using a secondary electron multiplier (SEM) or multiple-ion counters (MICs). Relative standard deviations (RSD) as small as 0.1% and abundance sensitivities of 1 x 10{sup 6} or better using a SEM are reported here. Precisions of RSD {approx} 0.01-0.001% using a multi-collector Faraday cup array can be achieved at sub-nanogram concentrations for strontium and neodymium and are suitable to gain crucial geolocation information. The analytical protocols reported herein are of particular value for nuclear forensic and nuclear safeguard applications.

  3. Development of a multi-layer diffusion couple to study fission product transport in ?-SiC

    NASA Astrophysics Data System (ADS)

    Dwaraknath, S.; Was, G. S.

    2014-01-01

    A multi-layer diffusion couple was designed to study fission product diffusion behavior while avoiding the pitfalls of direct ion implantation. Thin films of highly anisotropic pyrolytic carbon (PyC) were deposited onto CVD ?-SiC substrates. The PyC films were subsequently implanted with 400 keV silver, cesium, strontium, europium, or iodine at 22 C to a dose of 1016 cm-2, such that the implanted species resided wholly within the PyC layer. The samples were then coated with 50 nm of SiC via plasma enhanced CVD (PECVD) to retain the implanted species during post-deposition annealing treatments. The design allows for high spatial resolution tracking of the implanted specie using Rutherford backscattering spectrometry. Annealing at 1100 C for 10 h resulted in retention of 100% of implanted cesium, strontium, europium and iodine, and 70% of silver. This diffusion couple design provides the opportunity to determine diffusion coefficients of FPs in PyC and SiC and the role of the PyC/SiC interface in FP transport.

  4. Using tea as an artificial urine in a Canadian performance testing program for fission/activation products.

    PubMed

    Daka, Joseph N; Moodie, Gerry; DiNardo, Anthony; Kramer, Gary H

    2014-12-01

    In recent years, the National Calibration Reference Centre for Bioassay and In Vivo Monitoring (NCRC) at the Radiation Protection Bureau (RPB), Health Canada, has been conducting investigations with black tea to develop a matrix that can be used to replace urine in each of the following performance testing programs (PTP): (1) tritium, (2) carbon-14, (3) the DUAL (i.e., 3H/14C), and (4) fission/activation products (F/AP). A 1% tea solution with thimerosal, which had worked successfully for tritium, carbon-14, and the DUAL, was selected and tested for the F/AP PTP because of its similarity to urine in color and UV-VIS spectra. However, application of this tea to samples of the F/AP program containing 133Ba, 137Cs, 57Co, and 60Co produced precipitates, which was an unexpected result. Further experiments showed that replacement of thimerosal with an alcohol at about 5% eliminated the precipitation problem. The alcohol can be ethanol, methanol, or isopropanol. In the experiments, the 1% tea, preserved with alcohol, remained clear and stable for at least 100 d. The duration of each PTP for the NCRC is limited to 90 d. Application of the CNSC S-106 regulatory standard to the tea produced acceptable accuracy and precision results. It was concluded that a suitable tea matrix for the F/AP program had been found. PMID:25353238

  5. Thermodynamic limits to the quality of UCG product gas

    Microsoft Academic Search

    Creighton

    1982-01-01

    The goal of this work was to find the limits placed on the quality of UCG product gas by the energy and mass balances, including atom balances. If the outlet gas contains no O, there are only two independent variables. If these are chosen to be the mass fractions, X\\/sub CO\\/ and X\\/sub H\\/, both the temperature of the outlet

  6. Gas-to-electric furnace conversion hikes production

    Microsoft Academic Search

    J. Pare; G. Eklund

    1978-01-01

    The natural gas shortage in 1974 caused a manufacturer of motor gear trains to convert its carburizing furnace from gas to electric heating elements. The results of this retrofitting included a 40% increase in plant production, more variability in the use of the furnace for both low and high temperature applications, greater furnace loading capacity, reduced overall-operating costs, and elimination

  7. Trends in the short-term release of fission products and actinides to aqueous solution from used CANDU fuels at elevated temperature

    Microsoft Academic Search

    S. Stroes-Gascoyne

    1992-01-01

    A large number of short-term leaching experiments has been performed to determine fission product and actinide release from used CANDU (CANada Deuterium Uranium) fuels and to establish which factors affect release. Results are reported after30 10 d leaching at 100 150C under oxidizing (air) or reducing (Ar-3% H2 or Ar) conditions, in various synthetic groundwaters. Cesium-137 release (0.007 6%)

  8. Challenges, uncertainties and issues facing gas production from gas hydrate deposits

    SciTech Connect

    Moridis, G.J.; Collett, T.S.; Pooladi-Darvish, M.; Hancock, S.; Santamarina, C.; Boswell, R.; Kneafsey, T.; Rutqvist, J.; Kowalsky, M.; Reagan, M.T.; Sloan, E.D.; Sum, A.K.; Koh, C.

    2010-11-01

    The current paper complements the Moridis et al. (2009) review of the status of the effort toward commercial gas production from hydrates. We aim to describe the concept of the gas hydrate petroleum system, to discuss advances, requirement and suggested practices in gas hydrate (GH) prospecting and GH deposit characterization, and to review the associated technical, economic and environmental challenges and uncertainties, including: the accurate assessment of producible fractions of the GH resource, the development of methodologies for identifying suitable production targets, the sampling of hydrate-bearing sediments and sample analysis, the analysis and interpretation of geophysical surveys of GH reservoirs, well testing methods and interpretation of the results, geomechanical and reservoir/well stability concerns, well design, operation and installation, field operations and extending production beyond sand-dominated GH reservoirs, monitoring production and geomechanical stability, laboratory investigations, fundamental knowledge of hydrate behavior, the economics of commercial gas production from hydrates, and the associated environmental concerns.

  9. Preliminary report on the commercial viability of gas production from natural gas hydrates

    USGS Publications Warehouse

    Walsh, M.R.; Hancock, S.H.; Wilson, S.J.; Patil, S.L.; Moridis, G.J.; Boswell, R.; Collett, T.S.; Koh, C.A.; Sloan, E.D.

    2009-01-01

    Economic studies on simulated gas hydrate reservoirs have been compiled to estimate the price of natural gas that may lead to economically viable production from the most promising gas hydrate accumulations. As a first estimate, $CDN2005 12/Mscf is the lowest gas price that would allow economically viable production from gas hydrates in the absence of associated free gas, while an underlying gas deposit will reduce the viability price estimate to $CDN2005 7.50/Mscf. Results from a recent analysis of the simulated production of natural gas from marine hydrate deposits are also considered in this report; on an IROR basis, it is $US2008 3.50-4.00/Mscf more expensive to produce marine hydrates than conventional marine gas assuming the existence of sufficiently large marine hydrate accumulations. While these prices represent the best available estimates, the economic evaluation of a specific project is highly dependent on the producibility of the target zone, the amount of gas in place, the associated geologic and depositional environment, existing pipeline infrastructure, and local tariffs and taxes. ?? 2009 Elsevier B.V.

  10. Process for production desulfurized of synthesis gas

    DOEpatents

    Wolfenbarger, James K. (Torrance, CA); Najjar, Mitri S. (Wappingers Falls, NY)

    1993-01-01

    A process for the partial oxidation of a sulfur- and silicate-containing carbonaceous fuel to produce a synthesis gas with reduced sulfur content which comprises partially oxidizing said fuel at a temperature in the range of 1900.degree.-2600.degree. F. in the presence of a temperature moderator, an oxygen-containing gas and a sulfur capture additive which comprises a calcium-containing compound portion, a sodium-containing compound portion, and a fluoride-containing compound portion to produce a synthesis gas comprising H.sub.2 and CO with a reduced sulfur content and a molten slag which comprises (1) a sulfur-containing sodium-calcium-fluoride silicate phase; and (2) a sodium-calcium sulfide phase.

  11. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  12. The measurement of retained fission gas compositions and their isotopic distributions in an irradiated oxide fuel by inert gas fusion-mass spectrometric analysis

    Microsoft Academic Search

    S. D. Park; Y. S. Park; Y. K. Ha; K. Song

    2011-01-01

    In this work, an easy, fast and reliable measurement technique for the quantitative determination of retained fission gases\\u000a in an irradiated oxide fuel was developed. Many experiments were conducted to determine the optimum conditions for fusion\\u000a of an oxide fuel, for the quantitative collection and measurements of the released gases. Ion implantation technology was\\u000a applied to make a krypton or

  13. Tempest gas turbine extends EGT product line

    SciTech Connect

    Chellini, R.

    1995-07-01

    With the introduction of the 7.8 MW (mechanical output) Tempest gas turbine, ECT has extended the company`s line of its small industrial turbines. The new Tempest machine, featuring a 7.5 MW electric output and a 33% thermal efficiency, ranks above the company`s single-shaft Typhoon gas turbine, rated 3.2 and 4.9 MW, and the 6.3 MW Tornado gas turbine. All three machines are well-suited for use in combined heat and power (CHP) plants, as demonstrated by the fact that close to 50% of the 150 Typhoon units sold are for CHP applications. This experience has induced EGT, of Lincoln, England, to announce the introduction of the new gas turbine prior to completion of the testing program. The present single-shaft machine is expected to be used mainly for industrial trial cogeneration. This market segment, covering the needs of paper mills, hospitals, chemical plants, ceramic industry, etc., is a typical local market. Cogeneration plants are engineered according to local needs and have to be assisted by local organizations. For this reason, to efficiently cover the world market, EGT has selected a number of associates that will receive from Lincoln completely engineered machine packages and will engineer the cogeneration system according to custom requirements. These partners will also assist the customer and dispose locally of the spares required for maintenance operations.

  14. Benchmarking nuclear fission theory

    NASA Astrophysics Data System (ADS)

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; Talou, P.

    2015-07-01

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. The purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  15. Elemental Fluorine-18 Gas: Enhanced Production and Availability

    SciTech Connect

    VanBrocklin, Henry F. [Department of Radiology and Biomedical Imaging

    2011-12-01

    The overall objective of this project was to develop an efficient, reproducible and reliable process for the preparation of fluorine-18 labeled fluorine gas ([?F]F?) from readily available cyclotron-produced [?F]fluoride ion. The two step process entailed the production of [?F]fluoromethane with subsequent conversion to [?F]F? by electric discharge of [?F]fluoromethane in the presence of carrier nonradioactive F? gas. The specific goals of this project were i) to optimize the preparation of [?F]fluoromethane from [?F]fluoride ion; ii) to develop a prototype automated system for the production of [?F]F? from [?F]fluoride ion and iii) develop a compact user friendly automated system for the preparation of [?F]F? with initial synthesis of fluorine-18 labeled radiotracers. Over the last decade there has been an increased interest in the production of "non-standard" positron-emitting isotopes for the preparation of new radiotracers for a variety of applications including medical imaging and therapy. The increased availability of these isotopes from small biomedical cyclotrons has prompted their use in labeling radiotracers. In much the same way the production of [?F]F? gas has been known for several decades. However, access to [?F]F? gas has been limited to those laboratories with the means (e.g. F? targetry for the cyclotron) and the project-based need to work with [?F]F? gas. Relatively few laboratories, compared to those that produce [?F]fluoride ion on a daily basis, possess the capability to produce and use [?F]F? gas. A simplified, reliable system employing [?F]fluoride ion from cyclotron targetry systems that are already in place coupled with on-demand production of the [?F]F? gas would greatly enhance its availability. This would improve the availability of [?F]F? gas and promote further work with a valuable precursor. The major goals of the project were accomplished over the funding period. The preparation of ?F]fluoromethane has been automated with reproducible yields greater than 90% conversion from [?F]fluoride ion. A trap and release system was established for the [?F]fluoride ion concentration and direct elution of the [?F]fluoride ion into the reaction vial with the appropriate base and precursor in DMSO. Other solvents were also investigated. The production time for [?F]fluoromethane is less than 10 minutes. An automated system for the [?F]F? gas production from the [18F]fluoromethane has been developed. The unit coupled to the [?F]fluoromethane system permits the on demand production of [?F]F? gas. In less than 30 minutes, mCi quantities of [?F]F? gas were produced. Several variables for the [?F]F? gas production were investigated and a set of parameters for reproducible operation were determined. These parameters included discharge chamber size, carrier gas (He, Ne, Ar), discharge time, discharge current, mass of F? gas added to the chamber. FDOPA and EF5 were used to test the reactivity of the [?F]F? gas. Both products were produced in low to modest yield. The ready availability of [?F]F? gas has potential impact to advance both DOE mission-driven initiatives and nuclear medicine initiatives through other federally funded agencies such as NIH and DoD. New reactions involving the use of [?F]F? gas will lead to direct labeling of new radiotracers and intermediates as well as new fluorine-18 labeled synthons that may be further reacted with other reagents to provide useful fluorine-18 labeled compounds. New tracers to understand and follow plant and microbial metabolism as well as new tracers for nuclear medicine applications, that have been either difficult to obtain or never produced due to the limited availability of [?F]F? gas, may be prepared using the techniques developed .

  16. Evaluation of the gas production economics of the gas hydrate cyclic thermal injection model

    SciTech Connect

    Kuuskraa, V.A.; Hammersheimb, E.; Sawyer, W.

    1985-05-01

    The objective of the work performed under this directive is to assess whether gas hydrates could potentially be technically and economically recoverable. The technical potential and economics of recovering gas from a representative hydrate reservoir will be established using the cyclic thermal injection model, HYDMOD, appropriately modified for this effort, integrated with economics model for gas production on the North Slope of Alaska, and in the deep offshore Atlantic. The results from this effort are presented in this document. In Section 1, the engineering cost and financial analysis model used in performing the economic analysis of gas production from hydrates -- the Hydrates Gas Economics Model (HGEM) -- is described. Section 2 contains a users guide for HGEM. In Section 3, a preliminary economic assessment of the gas production economics of the gas hydrate cyclic thermal injection model is presented. Section 4 contains a summary critique of existing hydrate gas recovery models. Finally, Section 5 summarizes the model modification made to HYDMOD, the cyclic thermal injection model for hydrate gas recovery, in order to perform this analysis.

  17. Gas production by accelerated in situ bioleaching of landfills

    Microsoft Academic Search

    1982-01-01

    A process for improved gas production and accelerated stabilization of landfills by accelerated in situ bioleaching of organic wastes by acid forming bacteria in substantially sealed landfills, passing the leachate of hydrolysis and liquefaction products of microbial action of the microorganisms with the organic material to an acid phase digester to regenerate the activated culture of acid forming microorganisms for

  18. Methanol production from biomass and natural gas as transportation fuel

    Microsoft Academic Search

    Robert H. Borgwardt

    1998-01-01

    Two processes are examined for production of methanol. They are assessed against the essential requirements of a future alternative fuel for road transport: that it (1) is producible in amounts comparable to the 19 EJ of motor fuel annually consumed in the US, (2) minimizes emissions of criteria pollutants, (3) reduces greenhouse gas emissions from production and use, (4) is

  19. Chemical and Physical Properties of Dry Flue Gas Desulfurization Products

    Microsoft Academic Search

    David A. Kost; Jerry M. Bigham; Richard C. Stehouwer; Joel H. Beeghly; Randy Fowler; Samuel J. Traina; William E. Wolfe; Warren A. Dick

    2005-01-01

    be out of compliance without remedial action. This prob- lemhasspurredthedevelopmentofvarioustypesofscrub- Beneficial and environmentally safe recycling of flue gas desulfur- bing processes to convert SO2 from flue gases into solid ization (FGD) products requires detailed knowledge of their chemical and physical properties. We analyzed 59 dry FGD samples collected products for disposal or beneficial reuse. These FGD from 13 locations representing

  20. Natural gas productive capacity for the lower 48 states 1985 through 1997

    SciTech Connect

    NONE

    1996-12-01

    This publication presents information on wellhead productive capacity and a projection of gas production requirements. A history of natural gas production and productive capacity at the wellhead, along with a projection of the same, is illustrated.

  1. Fission dynamics within time-dependent Hartree-Fock: deformation-induced fission

    E-print Network

    Goddard, P M; Rios, A

    2015-01-01

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe fast fission processes beyond the fission barrier, using the nuclide $^{240}$Pu as an example. Methods: Time-dependent Hartree-Fock calculations based on the Skyrme interaction are used to calculate non-adiabatic fission paths, beginning from static constrained Hartree-Fock calculations. The properties of the dynamic states are interpreted in terms of the nature of their collective motion. Fission product properties are compared to data. Results: Parent nuclei constrained to begin dynamic evolution with a deformation less than the fission barrier exhibit giant-resonance-type behaviour. Those beginning just beyond the ...

  2. Helium production in natural gas reservoirs

    Microsoft Academic Search

    E. B. Pereira; J. A. S. Adams

    1982-01-01

    About 11,000 published natural gas analyses of helium are used in the estimation of the average global scale accumulation and concentration of radiogenic helium in sediments. Simple lognormal statistics is employed to derive a net accumulation rate between 1105 to 6.7105 helium atoms per cubic meter of reservoir rock per second. This acccumulation rate permitted to infer an average helium

  3. Helium production in natural gas reservoirs

    Microsoft Academic Search

    E. B. Pereira; J. A. S. Adams

    1982-01-01

    About 11,000 published natural gas analyses of helium are used in the estimation of the average global scale accumulation and concentration of radiogenic helium in sediments. Simple lognormal statistics is employed to derive a net accumulation rate between 1dagger10⁵ to 6.7dagger10⁵ helium atoms per cubic meter of reservoir rock per second. This acccumulation rate permitted to infer an average helium

  4. Bimodal fission

    SciTech Connect

    Hulet, E.K.

    1989-04-19

    In recent years, we have measured the mass and kinetic-energy distributions from the spontaneous fission of /sup 258/Fm, /sup 259/Md, /sup 260/Md, /sup 258/No, /sup 262/No, and /sup 260/(104). All are observed to fission with a symmetrical division of mass, whereas the total-kinetic-energy (TKE) distributions strongly deviated from the Gaussian shape characteristically found in the fission of all other actinides. When the TKE distributions are resolved into two Gaussians the constituent peaks lie near 200 and near 233 MeV. We conclude two modes or bimodal fission is occurring in five of the six nuclides studied. Both modes are possible in the same nuclides, but one generally predominates. We also conclude the low-energy but mass-symmetrical mode is likely to extend to far heavier nuclei; while the high-energy mode will be restricted to a smaller region, a region of nuclei defined by the proximity of the fragments to the strong neutron and proton shells in /sup 132/Sn. 16 refs., 7 figs., 1 tab.

  5. Canadian offshore oil production solution gas utilization alternatives

    SciTech Connect

    Wagner, J.V.

    1999-07-01

    Oil and gas development in the Province of Newfoundland and Labrador is in its early stage and the offshore industry emphasis is almost exclusively on oil production. At the Hibernia field, the Gravity Base Structure (GBS) is installed and the first wells are in production. The Terra Nova project, based on a Floating Production Storage Offloading (FPSO) ship shaped concept, is in its engineering and construction stage and first oil is expected by late 2000. Several other projects, such as Husky's White Rose and Chevron's Hebron, have significant potential for future development in the same area. It is highly probably that these projects will employ the FPSO concept. It is also expected that the solution gas disposal issues of such second generation projects will be of more significance in their regulatory approval process and of such second generation projects will be of more significance in their regulatory approval process and the operators may be forced to look for alternatives to gas reinjection. Three gas utilization alternatives for a FPSO concept based project have been considered and evaluated in this paper: liquefied natural gas (LNG), compressed natural gas (CNG), and gas-to-liquids conversion (GTL). The evaluation and the relative ranking of these alternatives is based on a first pass screening type of study which considers the technical and economical merits of each alternative. Publicly available information and in-house data, compiled within Fluor Daniel's various offices, was used to establish the basic parameters.

  6. Improving oil and gas production with the Beam Mounted Gas Compressor

    SciTech Connect

    AL-Khatib, A.M.

    1983-02-01

    This paper explores the parameters involved and the advantages obtained in the use of the BEAM MOUNTED GAS COMPRESSOR (B.M.G.C.). The B.M.G.C. is a single acting gas compressor operated by the walking beam of a rod pumping unit. The main function of the B.M.G.C. is to draw gas from the casing side of an oil well and discharge the gas into the flow line. By doing so, the B.M.G.C. operation actually reduces the back pressure on the formation face, thus allowing additional oil to enter the wellbore for production.

  7. Pumps, refracturing hike production from tight shale gas wells

    SciTech Connect

    Reeves, S.R. (Advanced Resources International Inc., Arlington, VA (United States)); Morrisson, W.K. (Nomeco Oil and Gas Co., Jackson, CO (United States)); Hill, D.G. (Gas Research Inst., Chicago, IL (United States))

    1993-02-01

    This paper reports that downhole pumps and refracturing are two ways to significantly improve production rates from the Antrim shale, a tight formation in the Michigan basin (U.S.) and the objective of a major natural gas play. Candidate wells for restimulation can be identified by pressure build-up tests and specifically productivity index-vs.-permeability plots based on these tests. The work in the Bagley East B4-10 well illustrates the possible production improvement.

  8. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. (eds.) (National Institute of Standards and Technology, Gaithersburg, MD (United States))

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  9. Production of Substitute Natural Gas from Coal

    SciTech Connect

    Andrew Lucero

    2009-01-31

    The goal of this research program was to develop and demonstrate a novel gasification technology to produce substitute natural gas (SNG) from coal. The technology relies on a continuous sequential processing method that differs substantially from the historic methanation or hydro-gasification processing technologies. The thermo-chemistry relies on all the same reactions, but the processing sequences are different. The proposed concept is appropriate for western sub-bituminous coals, which tend to be composed of about half fixed carbon and about half volatile matter (dry ash-free basis). In the most general terms the process requires four steps (1) separating the fixed carbon from the volatile matter (pyrolysis); (2) converting the volatile fraction into syngas (reforming); (3) reacting the syngas with heated carbon to make methane-rich fuel gas (methanation and hydro-gasification); and (4) generating process heat by combusting residual char (combustion). A key feature of this technology is that no oxygen plant is needed for char combustion.

  10. Automated production control in Arun gas Field

    SciTech Connect

    Scholz, W.; Aspoor, S.P.

    1984-02-01

    This paper describes the objectives, equipment and functions of the Supervisory Control and Data Acquisition (SCADA) system installed by Mobil Oil Indonesia (MOI) during the first half of 1983. The SCADA system provides operations, management, and engineering personnel with various applications of automatic production control such as production reports, field and cluster status, and flow calculations. This paper concludes with a discussion of the particular benefits gained by MOI in the Arun Field as well as the features of the SCADA system which make it a potential model for similar systems throughout the petroleum industry.

  11. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. (eds.) (National Institute of Standards and Technology, Gaithersburg, MD (United States))

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  12. Microscopic description of complex nuclear decay: Multimodal fission

    NASA Astrophysics Data System (ADS)

    Staszczak, A.; Baran, A.; Dobaczewski, J.; Nazarewicz, W.

    2009-07-01

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  13. Microscopic description of complex nuclear decay: multimodal fission

    E-print Network

    Staszczak, A; Dobaczewski, J; Nazarewicz, W

    2009-01-01

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  14. Microscopic description of complex nuclear decay: multimodal fission

    E-print Network

    A. Staszczak; A. Baran; J. Dobaczewski; W. Nazarewicz

    2009-06-23

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  15. Microscopic description of complex nuclear decay: Multimodal fission

    SciTech Connect

    Staszczak, A.; Baran, A. [Institute of Physics, Maria Curie-Sklodowska University, pl. M. Curie-Sklodowskiej 1, PL-20-031 Lublin (Poland); Department of Physics and Astronomy, University of Tennessee Knoxville, Tennessee 37996 (United States); Physics Division, Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, Tennessee 37831 (United States); Dobaczewski, J. [Institute of Theoretical Physics, University of Warsaw, ul. Hoza 69, PL-00-681 Warsaw (Poland); Department of Physics, P. O. Box 35 (YFL), FI-40014 University of Jyvaeskylae (Finland); Nazarewicz, W. [Department of Physics and Astronomy, University of Tennessee Knoxville, Tennessee 37996 (United States); Physics Division, Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, Tennessee 37831 (United States); Institute of Theoretical Physics, University of Warsaw, ul. Hoza 69, PL-00-681 Warsaw (Poland)

    2009-07-15

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  16. Background radiation from fission pulses

    SciTech Connect

    England, T.R.; Arthur, E.D.; Brady, M.C.; LaBauve, R.J.

    1988-05-01

    Extensive source terms for beta, gamma, and neutrons following fission pulses are presented in various tabular and graphical forms. Neutron results from a wide range of fissioning nuclides (42) are examined and detailed information is provided for four fuels: /sup 235/U, /sup 238/U, /sup 232/Th, and /sup 239/Pu; these bracket the range of the delayed spectra. Results at several cooling (decay) times are presented. For ..beta../sup -/ and ..gamma.. spectra, only /sup 235/U and /sup 239/Pu results are given; fission-product data are currently inadequate for other fuels. The data base consists of all known measured data for individual fission products extensively supplemented with nuclear model results. The process is evolutionary, and therefore, the current base is summarized in sufficient detail for users to judge its quality. Comparisons with recent delayed neutron experiments and total ..beta../sup -/ and ..gamma.. decay energies are included. 27 refs., 47 figs., 9 tabs.

  17. Gas chromatographic determination of the composition of coal pyridine products

    SciTech Connect

    Nabivach, V.M.; Berlizov, Yu.S.; Degtyareva, L.V.; Mariich, L.I.; Markus, G.A.

    1983-01-01

    The coking industry produces a wide range of products of the pyridine group from the crude light pyridine bases extracted from the coke oven gas. They include pure pyridine, pyridine solvent, ..cap alpha..-picoline, and the ..beta..-picoline and 2,4-lutidine fractions. The specifications for quality of these products are becoming increasingly stringent. The present methods of determining the quality indices are complex and laborious and do not always reflect the true component composition of the pyridine products. This paper describes the testing of a capillary gas chromatograph for the monitoring and analysis of pyridine products at all stages of their refining. The proposed method permits quantitative identification and adequately accurate determination of the concentraton of pyridine products. 15 references, 6 figures, 4 tables.

  18. Synthesis gas production by mixed conducting membranes with integrated conversion into liquid products

    DOEpatents

    Nataraj, Shankar (Allentown, PA); Russek, Steven Lee (Allentown, PA); Dyer, Paul Nigel (Allentown, PA)

    2000-01-01

    Natural gas or other methane-containing feed gas is converted to a C.sub.5 -C.sub.19 hydrocarbon liquid in an integrated system comprising an oxygenative synthesis gas generator, a non-oxygenative synthesis gas generator, and a hydrocarbon synthesis process such as the Fischer-Tropsch process. The oxygenative synthesis gas generator is a mixed conducting membrane reactor system and the non-oxygenative synthesis gas generator is preferably a heat exchange reformer wherein heat is provided by hot synthesis gas product from the mixed conducting membrane reactor system. Offgas and water from the Fischer-Tropsch process can be recycled to the synthesis gas generation system individually or in combination.

  19. Fission meter

    DOEpatents

    Rowland, Mark S. (Alamo, CA); Snyderman, Neal J. (Berkeley, CA)

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  20. Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model

    E-print Network

    Diecker, Jane T

    2005-01-01

    The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

  1. Some modern notions on oil and gas reservoir production regulation

    SciTech Connect

    Lohrenz, J.; Monash, E.A.

    1980-05-21

    The historic rhetoric of oil and gas reservoir production regulations has been burdened with misconceptions. One was that most reservoirs are rate insensitive. Another was that a reservoir's decline is primarily a function of reservoir mechaism rather than a choice unconstrained by the laws of physics. Relieved of old notions like these, we introduce some modern notions, the most basic being that production regulation should have the purpose of obtaining the highest value from production per irreversible diminution of thermodynamically available energy. The laws of thermodynamics determine the available energy. What then is value. Value may include contributions other than production per se and purely monetary economic outcomes.

  2. 40 CFR Table W - 1A of Subpart W-Default Whole Gas Emission Factors for Onshore Petroleum and Natural Gas Production

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...Factors for Onshore Petroleum and Natural Gas Production W Table W Protection...GREENHOUSE GAS REPORTING Petroleum and Natural Gas Systems Definitions. Pt...Factors for Onshore Petroleum and Natural Gas Production Onshore...

  3. Cascade heat recovery with coproduct gas production

    DOEpatents

    Brown, W.R.; Cassano, A.A.; Dunbobbin, B.R.; Rao, P.; Erickson, D.C.

    1986-10-14

    A process for the integration of a chemical absorption separation of oxygen and nitrogen from air with a combustion process is set forth wherein excess temperature availability from the combustion process is more effectively utilized to desorb oxygen product from the absorbent and then the sensible heat and absorption reaction heat is further utilized to produce a high temperature process stream. The oxygen may be utilized to enrich the combustion process wherein the high temperature heat for desorption is conducted in a heat exchange preferably performed with a pressure differential of less than 10 atmospheres which provides considerable flexibility in the heat exchange. 4 figs.

  4. Optimization and Evaluation of Mixed-Bed Chemisorbents for Extracting Fission and Activation Products from Marine and Fresh Waters

    SciTech Connect

    Johnson, Bryce; Santschi, Peter H.; Addleman, Raymond S.; Douglas, Matthew; Davidson, Joseph D.; Fryxell, Glen E.; Schwantes, Jon M.

    2011-06-02

    Chemically selective chemisorbents are needed to monitor natural and engineered waters for anthropogenic releases of stable and radioactive contaminants. Here, a number of individual and mixtures of chemisorbents were investigated for their ability to extract select fission and activation product elements from marine and coastal waters, including Co, Zr, Ru, Ag, Te, Sb, Ba, Cs, Ce, Eu, Pa, Np, and Th. Conventional manganese oxide and cyanoferrate sorbents, including commercially available Anfezh and potassium hexacyanocobalt(II) ferrate(II) (KCFC), were tested along with novel nano-structured surfaces (known as Self Assembled Monolayers on Mesoporous Supports or SAMMS) functionalized with a variety of moieties including thiol, diphosphonic acid (DiPhos-), methyl, 3, 4 hydroxypyridinone (HOPO-), and cyanoferrate. Extraction efficiencies were measured as a function of salinity, organic content, temperature, flow rate and sample size for both synthetic and natural fresh and saline waters under a range of environmentally relevant conditions. The effect of flow rate on extraction efficiency, from 1 to 70 mL min-1, provided some insight on rate limitations of mechanisms affecting sorption processes. Optimized mixtures of sorbent-ligand chemistries afforded excellent retention of all target elements, except, Ba and Sb. Mixtures of tested chemisorbents, including MnO2/Anfezh and MnO2/KCFC/Thiol (1-3mm)-SAMMS, extracted 8 of the 11 target elements studied to better than 80% efficiency, while a mixture of MnO2/Anfezh/Thiol (75-150 {mu}m)-SAMMS mixture was able to extract 7 of the 11 target elements to better than 90%. Results generated here indicate that flow rate should be less of a consideration for experimental design if sampling from fresh water containing variable amounts of DOM, rather than collecting samples from salt water environments. Relative to the capability of any single type of chemisorbent tested, optimized mixtures of several sorbents are able to increase the number of elements that can be efficiently and simultaneously extracted from natural waters.

  5. Reactive oxygen species production and discontinuous gas exchange in insects.

    PubMed

    Boardman, Leigh; Terblanche, John S; Hetz, Stefan K; Marais, Elrike; Chown, Steven L

    2012-03-01

    While biochemical mechanisms are typically used by animals to reduce oxidative damage, insects are suspected to employ a higher organizational level, discontinuous gas exchange mechanism to do so. Using a combination of real-time, flow-through respirometry and live-cell fluorescence microscopy, we show that spiracular control associated with the discontinuous gas exchange cycle (DGC) in Samia cynthia pupae is related to reactive oxygen species (ROS). Hyperoxia fails to increase mean ROS production, although minima are elevated above normoxic levels. Furthermore, a negative relationship between mean and mean ROS production indicates that higher ROS production is generally associated with lower . Our results, therefore, suggest a possible signalling role for ROS in DGC, rather than supporting the idea that DGC acts to reduce oxidative damage by regulating ROS production. PMID:21865257

  6. Bulk-nanocrystalline oxide nuclear fuels - An innovative material option for increasing fission gas retention, plasticity and radiation-tolerance

    NASA Astrophysics Data System (ADS)

    Spino, J.; Santa Cruz, H.; Jovani-Abril, R.; Birtcher, R.; Ferrero, C.

    2012-03-01

    Advantages and disadvantages of bulk nanocrystalline (nc)-oxides (UO2, ZrO2, ThO2) and suggestions for their potential use as nuclear fuels and inert matrix carriers are described in this work on the basis of a study with nc-4 mol% Y2O3-ZrO2 bodies, which are envisaged to behave akin to highly exposed LWR-fuels with the High Burn-up Structure (HBS) also known as rim transformation. The main attributes of nc-fuels in-pile compared to conventional fuels will be the capacity to develop closed porosity retaining most of the fission gases, the ability to relax more efficiently the interaction stresses with the cladding (through much higher plasticity) and the enhanced resistance against radiation-damage thanks to their nanostructure. The present analysis comprises the long-term thermal stability of a porous nc-material, its property vs. porosity relations, the topology of the pore phase via X-ray synchrotron tomography, the behaviour under compressive stress and the performance under intense Xe-ions irradiation. Salient outcomes are the non-connectivity of the pore phase, the superplasticity of the nc-bodies and their high radiation-amorphisation resistance with negligible swelling under Xe-bombardment. Another important outcome of the present study is that deterioration of the thermal properties due to grain boundary effects (Kapitza resistance, melting point depression) can likely be avoided if the grain size is kept above 100 nm and, emulating the real HBS material, preferably in the range between 200 and 300 nm.

  7. Fast-Mixed Spectrum Reactor progress report. Results of the FMSR Benchmark calculations and an assessment of current fission product libraries

    SciTech Connect

    Ludewig, H.; Durston, C.; Atefi, B.; Cerbone, R.J.

    1980-06-01

    As part of the Initial Feasibility Study of the Fast Mixed Spectrum Reactor, a series of benchmark calculations were made to determine the sensitivity of the physics analysis to differences in methods and data. Argonne National Laboratory (ANL), the Massachusetts Institute of Technology (MIT), and Oak Ridge National Laboratory (ORNL) were invited to participate with Brookhaven National Laboratory in the analysis of a FMSR model prescribed by BNL. Detailed comparisons are made including a comprehensive study on the adequacy of the fission product treatments.

  8. BUILDING MATERIALS MADE FROM FLUE GAS DESULFURIZATION BY-PRODUCTS

    SciTech Connect

    Michael W. Grutzeck; Maria DiCola; Paul Brenner

    2006-03-30

    Flue gas desulphurization (FGD) materials are produced in abundant quantities by coal burning utilities. Due to environmental restrains, flue gases must be ''cleaned'' prior to release to the atmosphere. They are two general methods to ''scrub'' flue gas: wet and dry. The choice of scrubbing material is often defined by the type of coal being burned, i.e. its composition. Scrubbing is traditionally carried out using a slurry of calcium containing material (slaked lime or calcium carbonate) that is made to contact exiting flue gas as either a spay injected into the gas or in a bubble tower. The calcium combined with the SO{sub 2} in the gas to form insoluble precipitates. Some plants have been using dry injection of these same materials or their own Class C fly ash to scrub. In either case the end product contains primarily hannebachite (CaSO{sub 3} {center_dot} 1/2H{sub 2}O) with smaller amounts of gypsum (CaSO{sub 4} {center_dot} 2H{sub 2}O). These materials have little commercial use. Experiments were carried out that were meant to explore the feasibility of using blends of hannebachite and fly ash mixed with concentrated sodium hydroxide to make masonry products. The results suggest that some of these mixtures could be used in place of conventional Portland cement based products such as retaining wall bricks and pavers.

  9. Low permeability gas reservoir production using large hydraulic fractures

    E-print Network

    Holditch, Stephen A

    1970-01-01

    , eatending two hundred feet past the cavity. The situation simulated was for production with a con- stant well bore pressure. Production rate and cumulative gas produced were monitored as functions of time. From these par- ameters an economic evaluation... TABLE OF CONTENTS Page ACKNOWLEDGEMENTS LIST OF TABLES LIST OF FIGURES INT ROD UC T ION PROCEDURE . THEORY RESULTS Effect of Stimulation on Flow Rate Effect of Well Bore Pressure Effect of Formation Permeability Effect of Stimulation...

  10. Entropy Production and Thermal Conductivity of A Dilute Gas

    E-print Network

    Yong-Jun Zhang

    2011-02-16

    It is known that the thermal conductivity of a dilute gas can be derived by using kinetic theory. We present here a new derivation by starting with two known entropy production principles: the steepest entropy ascent (SEA) principle and the maximum entropy production (MEP) principle. A remarkable feature of the new derivation is that it does not require the specification of the existence of the temperature gradient. The known result is reproduced in a similar form.

  11. Gas Mixtures and Ozone Production in an Electrical Discharge

    Microsoft Academic Search

    Thomas J. Manning; Jerry Hedden

    2001-01-01

    The quantitative production of ozone (O3) with N2, O2, and Ar gas mixtures in an atmospheric pressure corona discharge (CD) is investigated. A five-part model is presented that explores the discharge conditions needed for optimum ozone production. One part of the model is the well-known relationship that correlates the discharge's voltage, frequency, gap, dielectric material, etc with the generator's yield.

  12. Benz[a]anthracene Biotransformation and Production of Ring Fission Products by Sphingobium sp. Strain KK22

    PubMed Central

    Kunihiro, Marie; Ozeki, Yasuhiro; Nogi, Yuichi; Hamamura, Natsuko

    2013-01-01

    A soil bacterium, designated strain KK22, was isolated from a phenanthrene enrichment culture of a bacterial consortium that grew on diesel fuel, and it was found to biotransform the persistent environmental pollutant and high-molecular-weight polycyclic aromatic hydrocarbon (PAH) benz[a]anthracene. Nearly complete sequencing of the 16S rRNA gene of strain KK22 and phylogenetic analysis revealed that this organism is a new member of the genus Sphingobium. An 8-day time course study that consisted of whole-culture extractions followed by high-performance liquid chromatography (HPLC) analyses with fluorescence detection showed that 80 to 90% biodegradation of 2.5 mg liter?1 benz[a]anthracene had occurred. Biodegradation assays where benz[a]anthracene was supplied in crystalline form (100 mg liter?1) confirmed biodegradation and showed that strain KK22 cells precultured on glucose were equally capable of benz[a]anthracene biotransformation when precultured on glucose plus phenanthrene. Analyses of organic extracts from benz[a]anthracene biodegradation by liquid chromatography negative electrospray ionization tandem mass spectrometry [LC/ESI(?)-MS/MS] revealed 10 products, including two o-hydroxypolyaromatic acids and two hydroxy-naphthoic acids. 1-Hydroxy-2- and 2-hydroxy-3-naphthoic acids were unambiguously identified, and this indicated that oxidation of the benz[a]anthracene molecule occurred via both the linear kata and angular kata ends of the molecule. Other two- and single-aromatic-ring metabolites were also documented, including 3-(2-carboxyvinyl)naphthalene-2-carboxylic acid and salicylic acid, and the proposed pathways for benz[a]anthracene biotransformation by a bacterium were extended. PMID:23686261

  13. Environmental Compliance for Oil and Gas Exploration and Production

    SciTech Connect

    Hansen, Christine

    1999-10-26

    The Appalachian/Illinois Basin Directors is a group devoted to increasing communication among the state oil and gas regulatory agencies within the Appalachian and Illinois Basin producing region. The group is comprised of representatives from the oil and gas regulatory agencies from states in the basin (Attachment A). The directors met to discuss regulatory issues common to the area, organize workshops and seminars to meet the training needs of agencies dealing with the uniqueness of their producing region and perform other business pertinent to this area of oil and gas producing states. The emphasis of the coordinated work was a wide range of topics related to environmental compliance for natural gas and oil exploration and production.

  14. Dust and gas discharges in refractory production plants

    Microsoft Academic Search

    Yu. M. Svekrov; B. N. Maksimov; E. A. Orlova; A. M. Kasimov

    1983-01-01

    Conclusions The qualitative and quantitative characteristics of dust and gas discharges of the exhaust systems of production facilities in typical large plants of the refractory industry were determined and the values of the gross and specific discharges of harmful substances established.

  15. Production of synthesis gas and hydrogen from solid fuel

    Microsoft Academic Search

    D. Y. Gamburg; V. P. Semenov

    1983-01-01

    Conversion of the synthetic ammonia, methanol and hydrogen industry to solid fuel as a source of raw materials and energy will expand the raw material and energy base more than ten times compared to the use of natural hydrocarbon raw materials. This conversion assures stable future development in the production of synthetic liquid fuel (SLF) and synthetic natural gas (SNG).

  16. Devonian shale gas production; Mechanisms and simple models

    SciTech Connect

    Carlson, E.S. (Univ. of Alabama (US)); Mercer, J.C. (Dept. of Energy (US))

    1991-04-01

    This paper shows that, even without consideration of their special storage and flow properties, Devonian shales are special cases of dual porosity. The authors show that wile neglecting these properties in the short term is appropriate, such neglect in the long term will result in an under-estimation of shale gas production.

  17. Insights Into Natural Gas Production From Low-Permeability Reservoirs

    Microsoft Academic Search

    D. A. Northrop

    1988-01-01

    Insights have been gained into natural gas production from low permeability sandstone reservoirs in the western United States as a result of the US Department of Energy's Multiwell Experiment (MWX). Three wells, between 110 and 215 ft (34-66 m) apart at depth have been drilled at a site southwest of Rifle, Colorado, in the Piceance Basin, where the Cretaceous-age Mesaverde

  18. 78 FR 59632 - Oil and Gas and Sulphur Operations on the Outer Continental Shelf-Oil and Gas Production Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-27

    ...Continental Shelf--Oil and Gas Production Safety Systems AGENCY: Bureau of Safety...public comment period on the production safety systems proposed rule, which was...dry tree and subsea tree production systems on the Outer...

  19. Analytical Modeling of Shale Hydraulic Fracturing and Gas Production

    NASA Astrophysics Data System (ADS)

    Xu, W.

    2012-12-01

    Shale gas is abundant all over the world. Due to its extremely low permeability, extensive stimulation of a shale reservoir is always required for its economic production. Hydraulic fracturing has been the primary method of shale reservoir stimulation. Consequently the design and optimization of a hydraulic fracturing treatment plays a vital role insuring job success and economic production. Due to the many variables involved and the lack of a simple yet robust tool based on fundamental physics, horizontal well placement and fracturing job designs have to certain degree been a guessing game built on previous trial and error experience. This paper presents a method for hydraulic fracturing design and optimization in these environments. The growth of a complex hydraulic fracture network (HFN) during a fracturing job is equivalently represented by a wiremesh fracturing model (WFM) constructed on the basis of fracture mechanics and mass balance. The model also simulates proppant transport and placement during HFN growth. Results of WFM simulations can then be used as the input into a wiremesh production model (WPM) constructed based on WFM. WPM represents gas flow through the wiremesh HFN by an elliptic flow and the flow of gas in shale matrix by a novel analytical solution accounting for contributions from both free and adsorbed gases stored in the pore space. WPM simulation is validated by testing against numerical simulations using a commercially available reservoir production simulator. Due to the analytical nature of WFM and WPM, both hydraulic fracturing and gas production simulations run very fast on a regular personal computer and are suitable for hydraulic fracturing job design and optimization. A case study is presented to demonstrate how a non-optimized hydraulic fracturing job might have been optimized using WFM and WPM simulations.Fig. 1. Ellipsoidal representation of (a) stimulated reservoir and (b) hydraulic fracture network created by hydraulic fracturing treatment. Fig. 2. Gas flow represented by (a) elliptical flow through fracture network and (b) linear flow within reservoir matrix.

  20. 40 CFR Table Mm-1 to Subpart Mm of... - Default Factors for Petroleum Products and Natural Gas Liquids 1 2

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...Factors for Petroleum Products and Natural Gas Liquids 1 2 MM Table MM-1...Factors for Petroleum Products and Natural Gas Liquids 1 2 Products Column... Other Petroleum Products and Natural Gas Liquids Aviation Gasoline...

  1. 40 CFR Table Mm-1 to Subpart Mm of... - Default Factors for Petroleum Products and Natural Gas Liquids 1 2

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...Factors for Petroleum Products and Natural Gas Liquids 1 2 MM Table MM-1...Factors for Petroleum Products and Natural Gas Liquids 1 2 Products Column... Other Petroleum Products and Natural Gas Liquids Aviation Gasoline...

  2. 40 CFR Table Mm-1 to Subpart Mm of... - Default Factors for Petroleum Products and Natural Gas Liquids 1 2

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...Factors for Petroleum Products and Natural Gas Liquids 1 2 MM Table MM-1...Factors for Petroleum Products and Natural Gas Liquids 1 2 Products Column... Other Petroleum Products and Natural Gas Liquids Aviation Gasoline...

  3. 40 CFR Table Mm-1 to Subpart Mm of... - Default Factors for Petroleum Products and Natural Gas Liquids 1 2

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...Factors for Petroleum Products and Natural Gas Liquids 1 2 MM Table MM-1...Factors for Petroleum Products and Natural Gas Liquids 1 2 Products Column... Other Petroleum Products and Natural Gas Liquids Aviation Gasoline...

  4. Advanced Space Fission Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Borowski, Stanley K.

    2010-01-01

    Fission has been considered for in-space propulsion since the 1940s. Nuclear Thermal Propulsion (NTP) systems underwent extensive development from 1955-1973, completing 20 full power ground tests and achieving specific impulses nearly twice that of the best chemical propulsion systems. Space fission power systems (which may eventually enable Nuclear Electric Propulsion) have been flown in space by both the United States and the Former Soviet Union. Fission is the most developed and understood of the nuclear propulsion options (e.g. fission, fusion, antimatter, etc.), and fission has enjoyed tremendous terrestrial success for nearly 7 decades. Current space nuclear research and technology efforts are focused on devising and developing first generation systems that are safe, reliable and affordable. For propulsion, the focus is on nuclear thermal rockets that build on technologies and systems developed and tested under the Rover/NERVA and related programs from the Apollo era. NTP Affordability is achieved through use of previously developed fuels and materials, modern analytical techniques and test strategies, and development of a small engine for ground and flight technology demonstration. Initial NTP systems will be capable of achieving an Isp of 900 s at a relatively high thrust-to-weight ratio. The development and use of first generation space fission power and propulsion systems will provide new, game changing capabilities for NASA. In addition, development and use of these systems will provide the foundation for developing extremely advanced power and propulsion systems capable of routinely and affordably accessing any point in the solar system. The energy density of fissile fuel (8 x 10(exp 13) Joules/kg) is more than adequate for enabling extensive exploration and utilization of the solar system. For space fission propulsion systems, the key is converting the virtually unlimited energy of fission into thrust at the desired specific impulse and thrust-to-weight ratio. This presentation will discuss potential space fission propulsion options ranging from first generation systems to highly advanced systems. Ongoing research that shows promise for enabling second generation NTP systems with Isp greater than 1000 s will be discussed, as will the potential for liquid, gas, or plasma core systems. Space fission propulsion systems could also be used in conjunction with simple (water-based) propellant depots to enable routine, affordable missions to various destinations (e.g. moon, Mars, asteroids) once in-space infrastructure is sufficiently developed. As fuel and material technologies advance, very high performance Nuclear Electric Propulsion (NEP) systems may also become viable. These systems could enable sophisticated science missions, highly efficient cargo delivery, and human missions to numerous destinations. Commonalities between NTP, fission power systems, and NEP will be discussed.

  5. Mass distributions in monoenergetic-neutron-induced fission of 239Pu

    Microsoft Academic Search

    J. E. Gindler; L. E. Glendenin; D. J. Henderson; J. W. Meadows

    1983-01-01

    Fission product yields for 24 masses were determined for the fission of 239Pu with essentially monoenergetic neutrons of 0.17, 1.0, 2.0, 3.4, 4.5, 6.1, and 7.9 MeV. Fission product activities were measured by Ge(Li) gamma-ray spectrometry of irradiated 239Pu targets and by chemical separation of the fission product elements followed by beta counting. Yields of near symmetric (valley) fission products

  6. Gas plant economic optimization is more than meeting product specification

    SciTech Connect

    Berkowitz, P.N.; Colwell, L.W. [Continental Controls, Inc., Houston, TX (United States); Gamez, J.P. [Gas Research Inst., Chicago, IL (United States)

    1996-12-31

    Gas plants require a higher level of process control to optimize the process to maximize operating profits. Automation alone does not achieve this objective whereas, on-line dynamic optimization of the control variables based on product pricing, the cost to process the gas and the contracts for gas and liquids is solvable by new control techniques. Daily operations are affected by a paradigm shift in the method of control for the facility. This newly developed and site proven technique has demonstrated how to improve benefits when net processing margins are positive and minimize operating cost when liquids margins are negative. Because ethane recovery versus its rejection is not a binary decision, a better means to operate can be shown to benefit the gas plant operator. Each specification has a cost to meet it or a penalty to exceed it. However, if allowed, exceeding specification may prove beneficial to the net profitability of the operations. With the decision being made on-line every few minutes, the results are more dramatic than previously understood. Gas Research Institute and Continental Controls, Inc. have installed more than 10 such systems in US gas processing plants. Project payout from the use of the MVC{reg_sign} technology has on average been less than six months. Processing savings have ranged from $.0075 to $.024 per Mcf. The authors paper last year showed where the benefits can be derived. This year the results of those facilities are shared along with the methodology to achieve them.

  7. Fission dynamics within time-dependent Hartree-Fock: deformation-induced fission

    E-print Network

    P. M. Goddard; P. D. Stevenson; A. Rios

    2015-04-03

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe fast fission processes beyond the fission barrier, using the nuclide $^{240}$Pu as an example. Methods: Time-dependent Hartree-Fock calculations based on the Skyrme interaction are used to calculate non-adiabatic fission paths, beginning from static constrained Hartree-Fock calculations. The properties of the dynamic states are interpreted in terms of the nature of their collective motion. Fission product properties are compared to data. Results: Parent nuclei constrained to begin dynamic evolution with a deformation less than the fission barrier exhibit giant-resonance-type behaviour. Those beginning just beyond the barrier explore large amplitude motion but do not fission, whereas those beginning beyond the two-fragment pathway crossing fission to final states which differ according to the exact initial deformation. Conclusions: Time-dependent Hartree-Fock is able to give a good qualitative and quantitative description of fast fission, provided one begins from a sufficiently deformed state.

  8. NOBLE GAS PRODUCTION FROM MERCURY SPALLATION AT SNS

    SciTech Connect

    DeVore, Joe R [ORNL; Lu, Wei [ORNL; Schwahn, Scott O [ORNL

    2013-01-01

    Calculations for predicting the distribution of the products of spallation reactions between high energy protons and target materials are well developed and are used for design and operational applications in many projects both within DOE and in other arenas. These calculations are based on theory and limited experimental data that verifies rates of production of some spallation products exist. At the Spallation Neutron Source, a helium stream from the mercury target flows through a system to remove radioactivity from this mercury target offgas. The operation of this system offers a window through which the production of noble gases from mercury spallation by protons may be observed. This paper describes studies designed to measure the production rates of twelve noble gas isotopes within the Spallation Neutron Source mercury target.

  9. Trash-to-Gas: Converting Space Trash into Useful Products

    NASA Technical Reports Server (NTRS)

    Caraccio, Anne J.; Hintze, Paul E.

    2013-01-01

    NASA's Logistical Reduction and Repurposing (LRR) project is a collaborative effort in which NASA is determined to reduce total logistical mass through reduction, reuse and recycling of various wastes and components of long duration space missions and habitats. LRR is focusing on four distinct advanced areas of study: Advanced Clothing System, Logistics-to-Living, Heat Melt Compactor and Trash to Supply Gas (TtSG). The objective of TtSG is to develop technologies that convert material waste, human waste and food waste into high-value products. High-value products include life support oxygen and water, rocket fuels, raw material production feedstocks, and other energy sources. There are multiple pathways for converting waste to products involving single or multi-step processes. This paper discusses thermal oxidation methods of converting waste to methane. Different wastes, including food, food packaging, Maximum Absorbent Garments (MAGs), human waste simulants, and cotton washcloths have been evaluated in a thermal degradation reactor under conditions promoting pyrolysis, gasification or incineration. The goal was to evaluate the degradation processes at varying temperatures and ramp cycles and to maximize production of desirable products and minimize high molecular weight hydrocarbon (tar) production. Catalytic cracking was also evaluated to minimize tar production. The quantities of CO2, CO, CH4, and H2O were measured under the different thermal degradation conditions. The conversion efficiencies of these products were used to determine the best methods for producing desired products.

  10. Halogens in oil and gas production-associated wastewater.

    NASA Astrophysics Data System (ADS)

    Harkness, J.; Warner, N. R.; Dwyer, G. S.; Mitch, W.; Vengosh, A.

    2014-12-01

    Elevated chloride and bromide in oil and gas wastewaters that are released to the environment are one of the major environmental risks in areas impacted by shale gas development [Olmstead et al.,2013]. In addition to direct contamination of streams, the potential for formation of highly toxic disinfection by-products (DBPs) in drinking water in utilities located downstream from disposal sites poses a serious risk to human health. Here we report on the occurrence of iodide in oil and gas wastewater. We conducted systematic measurements of chloride, bromide, and iodide in (1) produced waters from conventional oil and gas wells from the Appalachian Basin; (2) hydraulic fracturing flowback fluids from unconventional Marcellus and Fayetteville shale gas, (3) effluents from a shale gas spill site in West Virginia; (4) effluents of oil and gas wastewater disposed to surface water from three brine treatment facilities in western Pennsylvania; and (5) surface waters downstream from the brine treatment facilities. Iodide concentration was measured by isotope dilution-inductively coupled plasma-mass spectrometry, which allowed for a more accurate measurement of iodide in a salt-rich matrix. Iodide in both conventional and unconventional oil and gas produced and flowback waters varied from 1 mg/L to 55 mg/L, with no systematic enrichment in hydraulic fracturing fluids. The similarity in iodide content between the unconventional Marcellus flowback waters and the conventional Appalachian produced waters clearly indicate that the hydraulic fracturing process does not induce additional iodide and the iodide content is related to natural variations in the host formations. Our data show that effluents from the brine treatment facilities have elevated iodide (mean = 20.91 mg/L) compared to local surface waters (0.03 0.1 mg/L). These results indicate that iodide, in addition to chloride and bromide in wastewater from oil and gas production, poses an additional risk to downstream surface water quality and drinking water utilities given the potential of formation of iodate-DBPs in drinking water. Olmstead, S.M. et al. (2013). Shale gas development impacts on surface water quality in Pennsylvania, PNAS, 110, 4962-4967.

  11. Fission gas release behaviour of a 103 GWd/tHM fuel disc during a 1200 C annealing test

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.; Tverberg, T.

    2014-03-01

    Within the Nuclear Fuel Industry Research (NFIR) program, several fuel variants, in the form of thin circular discs, were irradiated in the Halden Boiling Water Reactor (HBWR) to a range of burn-ups 100 GWd/tHM. The design of the assembly was similar to that used in other HBWR programs: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature gradients within the fuel discs. One such rod contained standard grain UO2 discs (3D grain size = 18 ?m) reaching a burn-up of 103 GWd/tHM. After the irradiation, the gas release upon rod puncturing was measured to be 2.9%.

  12. MANTRA: An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

    NASA Astrophysics Data System (ADS)

    Youinou, G.; Vondrasek, R.; Veselka, H.; Salvatores, M.; Paul, M.; Pardo, R.; Palmiotti, G.; Palchan, T.; Nusair, O.; Nimmagadda, J.; Nair, C.; Murray, P.; Maddock, T.; Kondrashev, S.; Kondev, F. G.; Jones, W.; Imel, G.; Glass, C.; Fonnesbeck, J.; Berg, J.; Bauder, W.

    2014-05-01

    This paper presents an update of an on-going collaborative INL-ANL-ISU integral reactor physics experiment whose objective is to infer the effective neutron capture cross sections for most of the actinides of importance for reactor physics and fuel cycle studies in both fast and epithermal spectra. Some fission products are also being considered. The principle of the experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation together with the neutron fluence will allow inference of effective neutron capture cross-sections in different neutron spectra.

  13. Metal Production in Quasars Through Jet-Gas Interactions

    NASA Astrophysics Data System (ADS)

    Vandegriff, Jon D.

    Emission lines studies of the gas surrounding many high redshift quasars indicate a high concentration of CNO nuclides. Relative abundance ratios may even exceed solar levels in some objects with redshifts near 5.0, indicating a rapid buildup of metals within one billion years after the big bang. Models explaining these high concentrations through standard stellar processing are pressed by the short time requirement. We explore a non-stellar nucleosynthesis mechanism in quasars based on the interaction of a high energy particle jet with hot, relatively dense gas. Although temperatures in the hot gas are high enough to support (thermalized) thermonuclear reactions, this mechanism alone is too slow to allow a rapid buildup of CNO nuclides. The collision of (non-thermal) jet particles with gas particles allows creation of unique nuclides which can boost the nucleosynthesis over traditional mass gaps at A = 5 and A = 8. The temperature and initial particle density range from T9=0.2 to T9=5.0, and 1011 to 1018 particles/cm3, respectively, while the jet intensity varies from 0.1 to 10 solar masses per year. The maximum final density allowed is 1023 particles/cm3. Substantial metal production in just 100 days can occur for temperatures near T9=0.6 and final densities of 1021 particles/cm3. Production at other temperatures and densities varies greatly. If the temperature is much above or below T9=0.6, or if the density cannot reach 1021 particles/cm3, then metal production is limited. Although the simple jet-clump model by itself does not seem capable of fully explaining the solar abundances in quasar gas, the low level production occurs on sufficiently short time scales so that it is still interesting. Also, a simplistic exploration of the production resulting from gas which evolves from high to low temperatures seems to indicate that at least 1/100th of solar levels can be obtained if the density can climb to 1021 particles/cm3 in a single processing episode of about 200 days. Multiple processing episodes and more complicated cooling scenarios may indicate larger nucleosynthesis possibilities. Therefore, the jet-clump model offers an exciting possibility for generating metals in quasars.

  14. Gas production by accelerated in situ bioleaching of landfills

    SciTech Connect

    Ghosh, S.

    1982-04-06

    A process for improved gas production and accelerated stabilization of landfills by accelerated in situ bioleaching of organic wastes by acid forming bacteria in substantially sealed landfills, passing the leachate of hydrolysis and liquefaction products of microbial action of the microorganisms with the organic material to an acid phase digester to regenerate the activated culture of acid forming microorganisms for recirculation to the landfill, passing the supernatant from the acid phase digester to a methane phase digester operated under conditions to produce methane rich gas. The supernatant from the methane phase digester containing nutrients for the acid forming microorganisms and added sewage sludge or other desired nutrient materials are circulated through the landfill. Low Btu gas is withdrawn from the acid phase digester while high Btu gas is withdrawn from the methane phase digester and may be upgraded for use as SNG. The process of this invention is applicable to small as well as large organic waste landfills, provides simultaneous disposal of municipal solid waste and sewage sludge or other aqueous organic waste in a landfill which may be stabilized much more quickly than an uncontrolled landfill as presently utilized.

  15. The fractal nature of the surface of uranium dioxide: a resolution of the short-lived\\/stable gas release dichotomy

    Microsoft Academic Search

    R. J. White

    2001-01-01

    The framework for analysis of fission product release in the Halden gas flow rigs has been developed over a period of 20 years. The predominant mode of release is single gas atom diffusion to free surfaces with a small, but burn-up-dependent contribution from direct recoil. Measurements of longer-lived fission products indicated that their release appeared to be controlled by a

  16. Production of synthesis gas in a solid electrolyte cell

    Microsoft Academic Search

    A. Kungolos; P. Tsiakaras; M. Stoukides

    1995-01-01

    The production of synthesis gas from methane was studied at 800-950C in an yttria-stabilized zirconia cell, using iron as\\u000a a catalyst-anode and platinum as cathodic electrode. The effect of gaseous O2 vs that of ionically transported O2- on CO selectivity and yield was studied. In general, O2- gave higher CO yields with maximum of 73%. The side reaction of hydrogen

  17. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  18. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  19. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  20. Production of bioplastics and hydrogen gas by photosynthetic microorganisms

    Microsoft Academic Search

    Asada Yasuo; Miyake Masato; Miyake Jun

    1998-01-01

    Our efforts have been aimed at the technological basis of photosynthetic-microbial production of materials and an energy carrier.\\u000a We report here accumulation of poly-(3-hydroxybutyrate) (PHB), a raw material of biodegradable plastics and for production\\u000a of hydrogen gas, and a renewable energy carrier by photosynthetic microorganisms (tentatively defined as cyanobacteria plus\\u000a photosynthetic bateria, in this report).\\u000a \\u000a A thermophilic cyanobacterium,Synechococcus sp. MA19

  1. Radiolytic gas production in the alpha particle degradation of plastics

    SciTech Connect

    Reed, D.T.; Hoh, J.; Emery, J. (Argonne National Lab., IL (United States)); Hobbs, D. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1992-01-01

    Net gas generation due to alpha particle irradiation of polyethylene and polyvinyl chloride was investigated. Experiments were performed in an air environment at 30, 60, and 100{degree}C. The predominant radiolytic degradation products of polyethylene were hydrogen and carbon dioxide with a wide variety of trace organic species noted. Irradiation of polyvinyl chloride resulted in the formation of HCl in addition to the products observed for polyethylene. For both plastic materials, a strong enhancement of net yields was noted at 100{degree}C.

  2. Radiolytic gas production in the alpha particle degradation of plastics

    SciTech Connect

    Reed, D.T.; Hoh, J.; Emery, J. [Argonne National Lab., IL (United States); Hobbs, D. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1992-05-01

    Net gas generation due to alpha particle irradiation of polyethylene and polyvinyl chloride was investigated. Experiments were performed in an air environment at 30, 60, and 100{degree}C. The predominant radiolytic degradation products of polyethylene were hydrogen and carbon dioxide with a wide variety of trace organic species noted. Irradiation of polyvinyl chloride resulted in the formation of HCl in addition to the products observed for polyethylene. For both plastic materials, a strong enhancement of net yields was noted at 100{degree}C.

  3. Measurement and calculation of the efficiency of fission detectors designed to monitor the time dependence of the neutron production of JET

    NASA Astrophysics Data System (ADS)

    Swinhoe, M. T.; Jarvis, O. N.

    1985-05-01

    Three pairs of fission counters (each pair one 235U and one 238U) are used at the Joint European Torus to determine the time dependence of the neutron production. In order to determine the absolute value of the neutron flux at the detector location it is necessary to know the neutron detection efficiency of the counter assemblies. This was measured using monoenergetic neutrons (at 2.5 and 14 MeV) and Cf and Am/Be sources. The fraction of fissions detected was determined by extrapolation of the pulse-height spectrum to zero pulse height. The calculation of efficiency was made with the Monte-Carlo neutron transport code MORSE. It was found that the detailed structure of the counter significantly affected the calculated efficiency and that the thermal cross-section values of the DLC37F nuclear data library had to be replaced with room-temperature values. The mean difference between calculation and experiment is (5.56.3)%.

  4. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ...2010-07-01 false How do I measure natural gas production on my eligible lease...Leases 260.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  5. Multiphasic analysis of gas production kinetics for in vitro fermentation of ruminant feeds

    Microsoft Academic Search

    Jeroen C. J. Groot; John W. Cone; Barbara A. Williams; Filip M. A. Debersaques; Egbert A. Lantinga

    1996-01-01

    Recently developed time-related gas production techniques to quantify the kinetics of ruminant feed fermentation have a high resolution. Consequently, fermentation processes with clearly contrasting gas production kinetics can be identified. Parameterization of the separate processes is possible with a suitable multiphasic model and modelling method. A flexible, empirical, multiphasic model was proposed for parameterization of gas production profiles. This equation

  6. Ranges in Air and Mass Identification of Plutonium Fission Fragments

    Microsoft Academic Search

    Seymour Katcoff; John A. Miskel; Charles W. Stanley

    1948-01-01

    Determinations were made of the mean and extrapolated ranges in air of plutonium fission fragments for twenty individual masses between 83 and 157. Collimated fission fragments passing through air at 120 or 140 mm pressure were deposited, after being stopped by the air, on a series of fourteen thin lacquer films. These were analyzed radio-chemically for individual fission products. The

  7. Production of bioplastics and hydrogen gas by photosynthetic microorganisms

    NASA Astrophysics Data System (ADS)

    Yasuo, Asada; Masato, Miyake; Jun, Miyake

    1998-03-01

    Our efforts have been aimed at the technological basis of photosynthetic-microbial production of materials and an energy carrier. We report here accumulation of poly-(3-hydroxybutyrate) (PHB), a raw material of biodegradable plastics and for production of hydrogen gas, and a renewable energy carrier by photosynthetic microorganisms (tentatively defined as cyanobacteria plus photosynthetic bateria, in this report). A thermophilic cyanobacterium, Synechococcus sp. MA19 that accumulates PHB at more than 20% of cell dry wt under nitrogen-starved conditions was isolated and microbiologically identified. The mechanism of PHB accumulation was studied. A mesophilic Synechococcus PCC7942 was transformed with the genes encoding PHB-synthesizing enzymes from Alcaligenes eutrophus. The transformant accumulated PHB under nitrogen-starved conditions. The optimal conditions for PHB accumulation by a photosynthetic bacterium grown on acetate were studied. Hydrogen production by photosynthetic microorganisms was studied. Cyanobacteria can produce hydrogen gas by nitrogenase or hydrogenase. Hydrogen production mediated by native hydrogenase in cyanobacteria was revealed to be in the dark anaerobic degradation of intracellular glycogen. A new system for light-dependent hydrogen production was targeted. In vitro and in vivo coupling of cyanobacterial ferredoxin with a heterologous hydrogenase was shown to produce hydrogen under light conditions. A trial for genetic trasformation of Synechococcus PCC7942 with the hydrogenase gene from Clostridium pasteurianum is going on. The strong hydrogen producers among photosynthetic bacteria were isolated and characterized. Co-culture of Rhodobacter and Clostriumdium was applied to produce hydrogen from glucose. Conversely in the case of cyanobacteria, genetic regulation of photosynthetic proteins was intended to improve conversion efficiency in hydrogen production by the photosynthetic bacterium, Rhodobacter sphaeroides RV. A mutant acquired by UV irradiation will be characterized for the mutation and for hydrogen productivity in comparison with the wild type strain. Some basic studies to develop photobioreactors are also introduced.

  8. Calculation of CO2 column heights in depleted gas fields from known pre-production gas column heights

    E-print Network

    Calculation of CO2 column heights in depleted gas fields from known pre-production gas column heights M. Naylor*, M. Wilkinson, R.S. Haszeldine School of GeoSciences, University of Edinburgh, EH9 3JW fields Column height a b s t r a c t Depleted gas fields have been identified as potential targets for CO

  9. Process for the production of fuel gas from coal

    DOEpatents

    Patel, Jitendra G. (Bolingbrook, IL); Sandstrom, William A. (Chicago, IL); Tarman, Paul B. (Elmhurst, IL)

    1982-01-01

    An improved apparatus and process for the conversion of hydrocarbonaceous materials, such as coal, to more valuable gaseous products in a fluidized bed gasification reaction and efficient withdrawal of agglomerated ash from the fluidized bed is disclosed. The improvements are obtained by introducing an oxygen containing gas into the bottom of the fluidized bed through a separate conduit positioned within the center of a nozzle adapted to agglomerate and withdraw the ash from the bottom of the fluidized bed. The conduit extends above the constricted center portion of the nozzle and preferably terminates within and does not extend from the nozzle. In addition to improving ash agglomeration and withdrawal, the present invention prevents sintering and clinkering of the ash in the fluidized bed and permits the efficient recycle of fine material recovered from the product gases by contacting the fines in the fluidized bed with the oxygen as it emanates from the conduit positioned within the withdrawal nozzle. Finally, the present method of oxygen introduction permits the efficient recycle of a portion of the product gases to the reaction zone to increase the reducing properties of the hot product gas.

  10. Mass distribution in 19F induced fission of 232Th

    Microsoft Academic Search

    G. K. Gubbi; A. Goswami; B. S. Tomar; B. John; A. Ramaswami; A. V. R. Reddy; P. P. Burte; S. B. Manohar

    1996-01-01

    Formation cross sections of several fission products have been determined using the recoil catcher technique followed by gamma-ray spectrometry in 19F induced fission of 232Th at Elab=95 and 112 MeV. The data show significant admixture of fission from compound nuclei formed by complete fusion as well as targetlike nuclei formed by transfer reactions. Mass distributions for both the fissioning systems

  11. Mass distribution in 11B induced fission of 232Th

    Microsoft Academic Search

    G. K. Gubbi; A. Goswami; B. S. Tomar; A. Ramaswami; A. V. R. Reddy; P. P. Burte; S. B. Manohar; B. John

    1999-01-01

    Formation cross sections of several fission products have been determined using recoil catcher technique followed by gamma-ray spectrometry in 11B induced fission of 232Th at Elab=72, 60, and 55 MeV. The data show significant admixture of fission from compound nucleus formed by complete fusion as well as targetlike nuclei formed by transfer reaction. Mass distributions for both the fissioning systems

  12. Characterizing tight-gas systems with production data: Wyoming, Utah, and Colorado

    USGS Publications Warehouse

    Nelson, Philip H.; Santus, Stephen L.

    2013-01-01

    The study of produced fluids allows comparisons among tight-gas systems. This paper examines gas, oil, and water production data from vertical wells in 23 fields in five Rocky Mountain basins of the United States, mostly from wells completed before the year 2000. Average daily rates of gas, oil, and water production are determined two years and seven years after production begins in order to represent the interval in which gas production declines exponentially. In addition to the daily rates, results are also presented in terms of oil-to-gas and water-to-gas ratios, and in terms of the five-year decline in gas production rates and water-to-gas ratios. No attempt has been made to estimate the ultimate productivity of wells or fields. The ratio of gas production rates after seven years to gas production rates at two years is about one-half, with median ratios falling within a range of 0.4 to 0.6 in 16 fields. Oil-gas ratios show substantial variation among fields, ranging from dry gas (no oil) to wet gas to retrograde conditions. Among wells within fields, the oil-gas ratios vary by a factor of three to thirty, with the exception of the Lance Formation in Jonah and Pinedale fields, where the oil-gas ratios vary by less than a factor of two. One field produces water-free gas and a large fraction of wells in two other fields produce water-free gas, but most fields have water-gas ratios greater than 1 bbl/mmcfgreater than can be attributed to water dissolved in gas in the reservoir and as high as 100 bbl/mmcf. The median water-gas ratio for fields increases moderately with time, but in individual wells water influx relative to gas is erratic, increasing greatly with time in many wells while remaining constant or decreasing in others.

  13. Production of noble gas isotopes by proton-induced reactions on bismuth

    NASA Astrophysics Data System (ADS)

    Leya, I.; David, J.-C.; Leray, S.; Wieler, R.; Michel, R.

    2008-04-01

    We measured integral thin target cross sections for the proton-induced production of He-, Ne-, Ar-, Kr- and Xe-isotopes from bismuth (Bi) from the respective reaction thresholds up to 2.6 GeV. Here we present 275 cross sections for 23 nuclear reactions. The production of noble gas isotopes from Bi is of special importance for design studies of accelerator driven systems (EA/ADS) and nuclear spallation sources. For experiments with proton energies above 200 MeV the mini-stack approach was used instead of the stacked-foil technique in order to minimise the influences of secondary particles on the residual nuclide production. Comparing the cross sections for Bi to the data published recently for Pb indicates that for 4He the cross sections for Bi below 200 MeV are up to a factor of 2-3 higher than the Pb data, which can be explained by the production of ?-decaying Po-isotopes from Bi but not from Pb. Some of the cross sections for the production of 21Ne from Bi are affected by recoil effects from neighboured Al-foils, which compromises a study of a possible lowering of the effective Coulomb-barrier. The differences in the excitation functions between Pb and Bi for Kr- and Xe-isotopes can be explained by energy-dependent higher fission cross sections for Bi compared to Pb. The experimental data are compared to results from the theoretical nuclear model codes INCL4/ABLA and TALYS. The INCL4/ABLA system describes the cross sections for the production of 4He-, Kr- and Xe-isotopes reasonably well, i.e. mostly within a factor of a few. In contrast, the model completely fails describing 21Ne, 22Ne, 36Ar and 38Ar, which are produced via spallation and/or multifragmentation. The TALYS code is only able to accurately predict reaction thresholds. The absolute values are either significantly over- or underestimated. Consequently, the comparison of measured and modelled thin target cross sections clearly indicates that experimental data are still needed because the predictive power of nuclear model codes, though permanently improving, does still not allow to reliably predict the cross sections needed for most applications and irradiation experiments remain indispensable.

  14. Ballistic piston fissioning plasma experiment.

    NASA Technical Reports Server (NTRS)

    Miller, B. E.; Schneider, R. T.; Thom, K.; Lalos, G. T.

    1971-01-01

    The production of fissioning uranium plasma samples such that the fission fragment stopping distance is less than the dimensions of the plasma is approached by using a ballistic piston device for the compression of uranium hexafluoride. The experimental apparatus is described. At room temperature the gun can be loaded up to 100 torr UF6 partial pressure, but at compression a thousand fold increase of pressure can be obtained at a particle density on the order of 10 to the 19th power per cu cm. Limited spectral studies of UF6 were performed while obtaining the pressure-volume data. The results obtained and their implications are discussed.

  15. Development of temporary subtropical wetlands induces higher gas production

    PubMed Central

    Canterle, Eliete B.; da Motta Marques, David; Rodrigues, Lcia R.

    2013-01-01

    Temporary wetlands are short-term alternative ecosystems formed by flooding for irrigation of areas used for rice farming. The goal of this study is to describe the development cycle of rice fields as temporary wetlands in southern Brazil, evaluating how this process affect the gas production (CH4 and CO2) in soil with difference % carbon and organic matter content. Two areas adjacent to Lake Mangueira in southern Brazil were used during a rice-farming cycle. One area had soil containing 1.1% carbon and 2.4% organic matter, and the second area had soil with 2.4% carbon and 4.4% organic matter. The mean rates of gas production were 0.04 0.02 mg CH4 m?2 d?1 and 1.18 0.30 mg CO2 m?2 d?1 in the soil area with the lower carbon content, and 0.02 0.03 mg CH4 m?2 d?1 and 1.38 0.41 mg CO2 m?2 d?1 in the soil area with higher carbon content. Our results showed that mean rates of CO2 production were higher than those of CH4 in both areas. No statistically significant difference was observed for production of CH4 considering different periods and sites. For carbon dioxide (CO2), however, a Two-Way ANOVA showed statistically significant difference (p = 0.05) considering sampling time, but no difference between areas. The results obtained suggest that the carbon and organic matter contents in the soil of irrigated rice cultivation areas may have been used in different ways by soil microorganisms, leading to variations in CH4 and CO2 production. PMID:23508352

  16. Trash to Gas: Converting Space Waste into Useful Supply Products

    NASA Technical Reports Server (NTRS)

    Tsoras, Alexandra

    2013-01-01

    The cost of sending mass into space with current propulsion technology is very expensive, making every item a crucial element of the space mission. It is essential that all materials be used to their fullest potential. Items like food, packaging, clothing, paper towels, gloves, etc., normally become trash and take up space after use. These waste materials are currently either burned up upon reentry in earth's atmosphere or sent on cargo return vehicles back to earth: a very wasteful method. The purpose of this project was to utilize these materials and create useful products like water and methane gas, which is used for rocket fuel, to further supply a deep space mission. The system used was a thermal degradation reactor with the configuration of a down-draft gasifier. The reactor was loaded with approximately 100g of trash simulant and heated with two external ceramic heaters with separate temperature control in order to create pyrolysis and gasification in one zone and incineration iri a second zone simultaneously. Trash was loaded into the top half of the reactor to undergo pyrolysis while the downdraft gas experienced gasification or incineration to treat tars and maximize the production of carbon dioxide. Minor products included carbon monoxide, methane, and other hydrocarbons. The carbon dioxide produced can be sent to a Sabatier reactor to convert the gas into methane, which can be used as rocket propellant. In order to maximize the carbon dioxide and useful gases produced, and minimize the unwanted tars and leftover ashen material, multiple experiments were performed with altered parameters such as differing temperatures, flow rates, and location of inlet air flow. According to the data received from these experiments, the process will be further scaled up and optimized to ultimately create a system that reduces trash buildup while at the same time providing enough useful gases to potentially fill a methane tank that could fuel a lunar ascent vehicle or other deep space mission.

  17. Chemical simulator for scale problems in oil and gas production

    SciTech Connect

    Morgnthaler, L.N.; Khatib, Z.I.; French, R.N.; Cox, K.R. (Shell Development Co., Houston, TX (United States))

    1991-04-01

    This paper reports that a chemical simulator incorporating state-of-the-art thermodynamics has been used to identify potential scale problems and design remedial treatments. In a West Texas waterflood, the success rate for gypsum removal treatments has more than doubled. Inhibitor squeezes are scheduled to prevent productivity losses from barium sulfate scale in an offshore waterflood. A freshwater injection program has been designed to prevent salt deposition in gas wells. In each case, the simulator played a key role in identifying potential scale problems and optimizing remedial measures.

  18. Thorium-uranium fission radiography

    NASA Technical Reports Server (NTRS)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  19. Agricultural use of a flue gas desulfurization by-product

    SciTech Connect

    Nelson, S. Jr.; Dick, W.; Chen, L.

    1998-07-01

    Few, if any, economical alternatives exist for operators of small coal-fired boilers that require a flue-gas desulfurization system which does not generate wastes. A new duct-injection technology called Fluesorbent has been developed to help fill this gap. Fluesorbent FGD was intentionally designed so that the saturated SO{sub 2}-sorbent materials would be valuable solid amendments for agricultural or turf-grass land. Agricultural and turf grass studies recently commenced using spent Fluesorbent materials from an FGD pilot program at an Ohio power plant. In the first year of testing, alfalfa yields on field plots with the FGS by-products were approximately 250% greater than on plots with no treatment, and about 40% greater than on plots treated with an equivalent amount of agricultural lime. Because the FGD by-products contained trace elements from included fly ash, the chemical composition of the alfalfa was significantly improved.

  20. Agricultural use of a flue gas desulfurization by-product

    SciTech Connect

    Nelson, S. Jr.; Dick, W.; Chen, L.

    1998-04-01

    Few, if any, economical alternatives exist for operators of small coal-fired boilers that require a flue-gas desulfurization system which does not generate wastes. A new duct-injection technology called {open_quotes}Fluesorbent{close_quotes} has been developed to help fill this gap. Fluesorbent FGD was intentionally designed so that the saturated SO{sub 2}-sorbent materials would be valuable soil amendments for agricultural or turf-grass land. Agricultural and turf grass studies recently commenced using spent Fluesorbent materials from an FGD pilot program at an Ohio power plant. In the first year of testing, alfalfa yields on field plots with the FGD by-products were approximately 250% greater than on plots with no treatment, and about 40% greater than on plots treated with an equivalent amount of agricultural lime. Because the FGD by-products contained trace elements from included fly ash, the chemical composition of the alfalfa was significantly improved.

  1. Air quality concerns of unconventional oil and natural gas production.

    PubMed

    Field, R A; Soltis, J; Murphy, S

    2014-05-01

    Increased use of hydraulic fracturing ("fracking") in unconventional oil and natural gas (O & NG) development from coal, sandstone, and shale deposits in the United States (US) has created environmental concerns over water and air quality impacts. In this perspective we focus on how the production of unconventional O & NG affects air quality. We pay particular attention to shale gas as this type of development has transformed natural gas production in the US and is set to become important in the rest of the world. A variety of potential emission sources can be spread over tens of thousands of acres of a production area and this complicates assessment of local and regional air quality impacts. We outline upstream activities including drilling, completion and production. After contrasting the context for development activities in the US and Europe we explore the use of inventories for determining air emissions. Location and scale of analysis is important, as O & NG production emissions in some US basins account for nearly 100% of the pollution burden, whereas in other basins these activities make up less than 10% of total air emissions. While emission inventories are beneficial to quantifying air emissions from a particular source category, they do have limitations when determining air quality impacts from a large area. Air monitoring is essential, not only to validate inventories, but also to measure impacts. We describe the use of measurements, including ground-based mobile monitoring, network stations, airborne, and satellite platforms for measuring air quality impacts. We identify nitrogen oxides, volatile organic compounds (VOC), ozone, hazardous air pollutants (HAP), and methane as pollutants of concern related to O & NG activities. These pollutants can contribute to air quality concerns and they may be regulated in ambient air, due to human health or climate forcing concerns. Close to well pads, emissions are concentrated and exposure to a wide range of pollutants is possible. Public health protection is improved when emissions are controlled and facilities are located away from where people live. Based on lessons learned in the US we outline an approach for future unconventional O & NG development that includes regulation, assessment and monitoring. PMID:24699994

  2. 79 FR 32502 - Managing Emissions From Oil and Natural Gas Production in Indian Country

    Federal Register 2010, 2011, 2012, 2013, 2014

    2014-06-05

    ...FRL-9910-71-OAR] RIN 2060-AS27 Managing Emissions From Oil and Natural Gas Production in Indian Country AGENCY...Minor New Source Review program for sources in the oil and natural gas production segment of the oil and natural gas sector. In particular,...

  3. Compact fission counter for DANCE

    SciTech Connect

    Wu, C Y; Chyzh, A; Kwan, E; Henderson, R; Gostic, J; Carter, D; Bredeweg, T; Couture, A; Jandel, M; Ullmann, J

    2010-11-06

    The Detector for Advanced Neutron Capture Experiments (DANCE) consists of 160 BF{sub 2} crystals with equal solid-angle coverage. DANCE is a 4{pi} {gamma}-ray calorimeter and designed to study the neutron-capture reactions on small quantities of radioactive and rare stable nuclei. These reactions are important for the radiochemistry applications and modeling the element production in stars. The recognition of capture event is made by the summed {gamma}-ray energy which is equivalent of the reaction Q-value and unique for a given capture reaction. For a selective group of actinides, where the neutron-induced fission reaction competes favorably with the neutron capture reaction, additional signature is needed to distinguish between fission and capture {gamma} rays for the DANCE measurement. This can be accomplished by introducing a detector system to tag fission fragments and thus establish a unique signature for the fission event. Once this system is implemented, one has the opportunity to study not only the capture but also fission reactions. A parallel-plate avalanche counter (PPAC) has many advantages for the detection of heavy charged particles such as fission fragments. These include fast timing, resistance to radiation damage, and tolerance of high counting rate. A PPAC also can be tuned to be insensitive to {alpha} particles, which is important for experiments with {alpha}-emitting actinides. Therefore, a PPAC is an ideal detector for experiments requiring a fast and clean trigger for fission. A PPAC with an ingenious design was fabricated in 2006 by integrating amplifiers into the target assembly. However, this counter was proved to be unsuitable for this application because of issues related to the stability of amplifiers and the ability to separate fission fragments from {alpha}'s. Therefore, a new design is needed. A LLNL proposal to develop a new PPAC for DANCE was funded by NA22 in FY09. The design goal is to minimize the mass for the proposed counter and still be able to maintain a stable operation under extreme radioactivity and the ability to separate fission fragments from {alpha}'s. In the following sections, the description is given for the design and performance of this new compact PPAC, for studying the neutron-induced reactions on actinides using DANCE at LANL.

  4. Noble gas isotope measurements for spent nuclear fuel reprocessing. IAEA Task 90/0A211 interim report

    SciTech Connect

    Hudson, G.B.

    1993-02-17

    The nuclear fission of actinides in reactor fuel produces large quantities of Kr and Xe as fission products. Because of the high levels of fission Kr and Xe, sample collection and analysis of noble gases for spent fuel diagnostic measurements is a simple, straight-forward technique. In modern reprocessing plants with continuous dissolvers, it will not be possible to use traditional methods for isolating input batches of fuel. This study investigates the feasibility of using noble gas isotope abundance measurements (isotope correlation techniques - ICT) to solve safeguards requirements. Noble gas measurements might be able to provide an independent analysis of Pu contained within dissolves fuel, on an individual fuel assembly basis. The isotopic composition of Kr and Xe in spent fuel reflects both the composition (isotope abundance ratios) of the fission products and the effects of neutron capture on those fission products. We have reviewed the available literature for noble gas analyses of spent reactor fuel. While references are made to noble gas isotope correlations over the last 20 years, we have found little if any detailed analysis of large data sets. The literature search did find several useful reports. Of these papers, one is particularly useful for evaluating noble gas isotopic compositions. The ``Benchmark-paper`` (1) contains 54 Kr and 56 Xe isotopic composition analyses for 4 different reactors with a variety of fuel enrichment factors. Burnup ranges from 8000 to 37000 MWd/tU. Besides the noble gas measurements, a variety of other measurements are reported (actinides and fission products).

  5. Event-by-Event Fission with FREYA

    SciTech Connect

    Randrup, J; Vogt, R

    2010-11-09

    The recently developed code FREYA (Fission Reaction Event Yield Algorithm) generates large samples of complete fission events, consisting of two receding product nuclei as well as a number of neutrons and photons, all with complete kinematic information. Thus it is possible to calculate arbitrary correlation observables whose behavior may provide unique insight into the fission process. The presentation first discusses the present status of FREYA, which has now been extended up to energies where pre-equilibrium emission becomes significant and one or more neutrons may be emitted prior to fission. Concentrating on {sup 239}Pu(n,f), we discuss the neutron multiplicity correlations, the dependence of the neutron energy spectrum on the neutron multiplicity, and the relationship between the fragment kinetic energy and the number of neutrons and their energies. We also briefly suggest novel fission observables that could be measured with modern detectors.

  6. Microbiology of synthesis gas fermentation for biofuel production.

    PubMed

    Henstra, Anne M; Sipma, Jan; Rinzema, Arjen; Stams, Alfons J M

    2007-06-01

    A significant portion of biomass sources like straw and wood is poorly degradable and cannot be converted to biofuels by microorganisms. The gasification of this waste material to produce synthesis gas (or syngas) could offer a solution to this problem, as microorganisms that convert CO and H2) (the essential components of syngas) to multicarbon compounds are available. These are predominantly mesophilic microorganisms that produce short-chain fatty acids and alcohols from CO and H2. Additionally, hydrogen can be produced by carboxydotrophic hydrogenogenic bacteria that convert CO and H2O to H2 and CO2. The production of ethanol through syngas fermentation is already available as a commercial process. The use of thermophilic microorganisms for these processes could offer some advantages; however, to date, few thermophiles are known that grow well on syngas and produce organic compounds. The identification of new isolates that would broaden the product range of syngas fermentations is desirable. Metabolic engineering could be employed to broaden the variety of available products, although genetic tools for such engineering are currently unavailable. Nevertheless, syngas fermenting microorganisms possess advantageous characteristics for biofuel production and hold potential for future engineering efforts. PMID:17399976

  7. Theoretical Description of the Fission Process

    SciTech Connect

    Witold Nazarewicz

    2009-10-25

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nations nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic waste and be proliferation-resistant, is a goal for the advanced nuclear fuel cycles program. While in the past the design, construction, and operation of reactors were supported through empirical trials, this new phase in nuclear energy production is expected to heavily rely on advanced modeling and simulation capabilities.

  8. Potential energy and greenhouse gas emission effects of hydrogen production from coke oven gas in U.S. steel mills

    Microsoft Academic Search

    Fred Joseck; Michael Wang; Ye Wu

    2008-01-01

    For this study, we examined the energy and emission effects of hydrogen production from coke oven gas (COG) on a well-to-wheels basis and compared these effects with those of other hydrogen production options, as well as with those of conventional gasoline and diesel options. We then estimated the magnitude of hydrogen production from COG in the United States and the

  9. Atmospheric emissions and air quality impacts from natural gas production and use.

    PubMed

    Allen, David T

    2014-01-01

    The US Energy Information Administration projects that hydraulic fracturing of shale formations will become a dominant source of domestic natural gas supply over the next several decades, transforming the energy landscape in the United States. However, the environmental impacts associated with fracking for shale gas have made it controversial. This review examines emissions and impacts of air pollutants associated with shale gas production and use. Emissions and impacts of greenhouse gases, photochemically active air pollutants, and toxic air pollutants are described. In addition to the direct atmospheric impacts of expanded natural gas production, indirect effects are also described. Widespread availability of shale gas can drive down natural gas prices, which, in turn, can impact the use patterns for natural gas. Natural gas production and use in electricity generation are used as a case study for examining these indirect consequences of expanded natural gas availability. PMID:24498952

  10. An evaluation of hydrogen production from the perspective of using blast furnace gas and coke oven gas as feedstocks

    Microsoft Academic Search

    Wei-Hsin Chen; Mu-Rong Lin; Tzong-Shyng Leu; Shan-Wen Du

    2011-01-01

    Blast furnace (BF) is a large-scale reactor for producing hot metal where coke and coal are consumed as reducing agent and fuel, respectively. As a result, a large amount of CO2 is liberated into the atmosphere. The blast furnace gas (BFG) and coke oven gas (COG) from the ironmaking process can be used for H2 production in association with carbon

  11. Evaluation of the gas production economics of the gas hydrate cyclic thermal injection model. [Cyclic thermal injection

    SciTech Connect

    Kuuskraa, V.A.; Hammersheimb, E.; Sawyer, W.

    1985-05-01

    The objective of the work performed under this directive is to assess whether gas hydrates could potentially be technically and economically recoverable. The technical potential and economics of recovering gas from a representative hydrate reservoir will be established using the cyclic thermal injection model, HYDMOD, appropriately modified for this effort, integrated with economics model for gas production on the North Slope of Alaska, and in the deep offshore Atlantic. The results from this effort are presented in this document. In Section 1, the engineering cost and financial analysis model used in performing the economic analysis of gas production from hydrates -- the Hydrates Gas Economics Model (HGEM) -- is described. Section 2 contains a users guide for HGEM. In Section 3, a preliminary economic assessment of the gas production economics of the gas hydrate cyclic thermal injection model is presented. Section 4 contains a summary critique of existing hydrate gas recovery models. Finally, Section 5 summarizes the model modification made to HYDMOD, the cyclic thermal injection model for hydrate gas recovery, in order to perform this analysis.

  12. Gas chromatographic determination of fenvalerate insecticide residues in processed tomato products and by-products.

    PubMed

    Spittler, T D; Argauer, R J; Lisk, D J; Mumma, R O; Winnett, G; Ferro, D N; Bogus, E; Greco, E; Gutenmann, W; Miles, E

    1984-01-01

    The results of a 5-laboratory collaborative determination of residues of the synthetic pyrethroid insecticide fenvalerate in tomato products are presented. Tomatoes from plants treated in the field at 2-4 day intervals (13 foliar applications) were processed into chopped fresh tomatoes, canned quarters, juice, paste, and the by-product skins plus seeds. Gas chromatographic analysis of the commodities for fenvalerate showed the fresh produce to contain 0.26 ppm, and the skins plus seeds contained 1.9 ppm. Residues were barely detectable in canned peeled quarters and juice, but averaged 0.12 ppm for paste, the concentration product of juice. High residues were associated with the skin content of the product. Five laboratories using modifications of the same analytical technique obtained good collaborative agreement. PMID:6469916

  13. Fifty years with nuclear fission. Volume 2

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. [eds.] [National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1989-12-31

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, ``Fifty years with nuclear fission,`` in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  14. Mass resolved angular distribution in 10B, 12C, and 16O induced fission of 232Th

    Microsoft Academic Search

    Bency John; Aruna Nijasure; S. K. Kataria; A. Goswami; B. S. Tomar; A. V. R. Reddy; S. B. Manohar

    1995-01-01

    The recoil catcher technique and gamma spectrometric assay of fission products were used to measure angular distribution of 17 fission products in 10B, 12C, and 16O induced fission of 232Th at near barrier energies. The observed mass dependence of anisotropies of fission products in these systems are very different from that of p and alpha induced fission of 232Th at

  15. Fifty years with nuclear fission. Volume 1

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. [eds.] [National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1989-12-31

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, ``Fifty Years with Nuclear Fission,`` in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  16. Fission of 232Th at energies up to 90 MeV

    Microsoft Academic Search

    Chien Chung; James J. Hogan

    1981-01-01

    In a recent work, the yields of spallation residues from 8-90 MeV proton induced fission of 232Th were reproduced by exciton model calculations to which a fission option had been added. This allowed evaluation of the spectrum of fissioning nuclei. In this work, an extension of that calculation to the properties of the fission products is presented. Dividing the fissioning

  17. Oil production by gas drive from adjacent strata

    Microsoft Academic Search

    Cornelius

    1969-01-01

    Oil is produced from an oil stratum lying adjacent but separated from a gas formation by an impermeable barrier by injecting an aqueous flood, such as water free of or containing additives, into the gas formation to force gas into a well communicating with the oil stratum, so as to force the gas into the oil stratum and drive oil

  18. Modeling the Relative GHG Emissions of Conventional and Shale Gas Production

    PubMed Central

    2011-01-01

    Recent reports show growing reserves of unconventional gas are available and that there is an appetite from policy makers, industry, and others to better understand the GHG impact of exploiting reserves such as shale gas. There is little publicly available data comparing unconventional and conventional gas production. Existing studies rely on national inventories, but it is not generally possible to separate emissions from unconventional and conventional sources within these totals. Even if unconventional and conventional sites had been listed separately, it would not be possible to eliminate site-specific factors to compare gas production methods on an equal footing. To address this difficulty, the emissions of gas production have instead been modeled. In this way, parameters common to both methods of production can be held constant, while allowing those parameters which differentiate unconventional gas and conventional gas production to vary. The results are placed into the context of power generation, to give a ?well-to-wire? (WtW) intensity. It was estimated that shale gas typically has a WtW emissions intensity about 1.82.4% higher than conventional gas, arising mainly from higher methane releases in well completion. Even using extreme assumptions, it was found that WtW emissions from shale gas need be no more than 15% higher than conventional gas if flaring or recovery measures are used. In all cases considered, the WtW emissions of shale gas powergen are significantly lower than those of coal. PMID:22085088

  19. Modeling the relative GHG emissions of conventional and shale gas production.

    PubMed

    Stephenson, Trevor; Valle, Jose Eduardo; Riera-Palou, Xavier

    2011-12-15

    Recent reports show growing reserves of unconventional gas are available and that there is an appetite from policy makers, industry, and others to better understand the GHG impact of exploiting reserves such as shale gas. There is little publicly available data comparing unconventional and conventional gas production. Existing studies rely on national inventories, but it is not generally possible to separate emissions from unconventional and conventional sources within these totals. Even if unconventional and conventional sites had been listed separately, it would not be possible to eliminate site-specific factors to compare gas production methods on an equal footing. To address this difficulty, the emissions of gas production have instead been modeled. In this way, parameters common to both methods of production can be held constant, while allowing those parameters which differentiate unconventional gas and conventional gas production to vary. The results are placed into the context of power generation, to give a ?well-to-wire? (WtW) intensity. It was estimated that shale gas typically has a WtW emissions intensity about 1.8-2.4% higher than conventional gas, arising mainly from higher methane releases in well completion. Even using extreme assumptions, it was found that WtW emissions from shale gas need be no more than 15% higher than conventional gas if flaring or recovery measures are used. In all cases considered, the WtW emissions of shale gas powergen are significantly lower than those of coal. PMID:22085088

  20. Industrial emergy evaluation for hydrogen production systems from biomass and natural gas

    Microsoft Academic Search

    Xiao Feng; Li Wang; Shuling Min

    2009-01-01

    Fossil fuel resources are the main source for hydrogen production, and hydrogen production by renewable energy, such as biomass, is under development. To compare the performance in natural resource utilization for different hydrogen production systems, in this paper, two laboratorial hydrogen production systems from biomass and one industrial hydrogen production system from natural gas are analyzed by using industrial emergy