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Sample records for gas fission products

  1. Fission gas detection system

    DOEpatents

    Colburn, Richard P.

    1985-01-01

    A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

  2. Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

    SciTech Connect

    B.R. Westphal; J.J. Park; J.M. Shin; G.I. Park; K.J. Bateman; D.L. Wahlquist

    2008-07-01

    A head-end processing step, termed DEOX for its emphasis on decladding via oxidation, is being developed for the treatment of spent oxide fuel by pyroprocessing techniques. The head-end step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Development of the head-end step is being performed in collaboration with the Korean Atomic Energy Research Institute (KAERI) through an International Nuclear Energy Research Initiative. Following the initial experimentation for the removal of volatile fission products, an off-gas treatment system was designed in conjunction with KAERI to collect specific fission gases. The primary volatile species targeted for trapping were iodine, technetium, and cesium. Each species is intended to be collected in distinct zones of the off-gas system and within those zones, on individual filters. Separation of the volatile off-gases is achieved thermally as well as chemically given the composition of the filter media. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. In addition, results from testing with the off-gas treatment system will be presented.

  3. 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.; Krämer, J.; Krieger, A.; Knuth, K.; Nagy, Sz.; Nörtershäuser, 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 γ-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.

  4. 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.

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

    SciTech Connect

    Dawn M. Scates; John 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/B’s) 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.

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

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    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/B’s) 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.

  7. Gas emission from the UO2 samples, containing fission products and burnable absorber

    NASA Astrophysics Data System (ADS)

    Kopytin, V. P.; Baranov, V. G.; Burlakova, M. A.; Tenishev, A. V.; Kuzmin, R. S.; Pokrovskiy, S. A.; Mikhalchik, V. V.

    2016-04-01

    The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

  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. 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.; Günther-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.

  10. Fission-product retention in HTGR fuels

    SciTech Connect

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

  11. 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.

  12. TREATMENT OF FISSION PRODUCT WASTE

    DOEpatents

    Huff, J.B.

    1959-07-28

    A pyrogenic method of separating nuclear reactor waste solutions containing aluminum and fission products as buring petroleum coke in an underground retort, collecting the easily volatile gases resulting as the first fraction, he uminum chloride as the second fraction, permitting the coke bed to cool and ll contain all the longest lived radioactive fission products in greatly reduced volume.

  13. An analytical study of volatile metallic fission product release from very high temperature gas-cooled reactor fuel and core

    SciTech Connect

    Mitake, S.; Okamoto, F.

    1988-04-01

    Release characteristics of volatile metallic fission products from the coated fuel particle and the reactor core for a very high temperature gas-cooled reactor during its power operation has been studied using numerical analysis. A computer code FORNAX, based on Fick's diffusion law and the evaporation mass transfer relation, has been developed, which considers, in particular, distribution and time histories of power density, fuel temperature, and failed and degraded fuel particle fractions in the core. Applicability of the code to evaluate the core design has been shown and the following have been indicated on the release of cesium from the reactor: 1. The release from the intact fuel particles by diffusion through their intact coatings shows larger contribution in the total core release at higher temperature. 2. The diffusion release from the intact particle is governed not only by the diffusion in the silicon carbide layer but also by that in the fuel kernel.

  14. Correlation of recent fission product release data

    SciTech Connect

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab.

  15. PRODUCING ENERGY AND RADIOACTIVE FISSION PRODUCTS

    DOEpatents

    Segre, E.; Kennedy, J.W.; Seaborg, G.T.

    1959-10-13

    This patent broadly discloses the production of plutonium by the neutron bombardment of uranium to produce neptunium which decays to plutonium, and the fissionability of plutonium by neutrons, both fast and thermal, to produce energy and fission products.

  16. 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.

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

    DOE PAGESBeta

    Meierbachtol, K.; Tovesson, F.; Shields, D.; Arnold, C.; Blakeley, R.; Bredeweg, T.; Devlin, M.; Hecht, A. A.; Heffern, L. E.; Jorgenson, J.; et al

    2015-04-01

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) 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 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, assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Finally,more » 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). These mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.« less

  18. 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.

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

    SciTech Connect

    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.

    2015-04-01

    The SPectrometer for Ion DEtermination in fission Research (SPIDER) 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 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, assembled and tested using 229Th and 252Cf radioactive decay sources. For commissioning, the fully assembled system measured fission products from spontaneous fission of 252Cf. Finally, 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). These mass yield results measured from 252Cf spontaneous fission products are reported from an E–v measurement.

  20. Antiproton Powered Gas Core Fission Rocket

    NASA Astrophysics Data System (ADS)

    Kammash, T.

    Extensive research in recent years has demonstrated that “at rest” annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  1. Antiproton Powered Gas Core Fission Rocket

    NASA Astrophysics Data System (ADS)

    Kammash, Terry

    2005-02-01

    Extensive research in recent years has demonstrated that "at rest" annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas — inserted into the chamber just prior to the release of the antiproton — to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  2. Antiproton Powered Gas Core Fission Rocket

    SciTech Connect

    Kammash, Terry

    2005-02-06

    Extensive research in recent years has demonstrated that 'at rest' annihilation of antiprotons in the uranium isotope U238 leads to fission at nearly 100% efficiency. The resulting highly-ionizing, energetic fission fragments can heat a suitable medium to very high temperatures, making such a process particularly suitable for space propulsion applications. Such an ionized medium, which would serve as a propellant, can be confined by a magnetic field during the heating process, and subsequently ejected through a magnetic nozzle to generate thrust. The gasdynamic mirror (GDM) magnetic configuration is especially suited for this application since the underlying confinement principle is that the plasma be of such density and temperature as to make the ion-ion collision mean free path shorter than the plasma length. Under these conditions the plasma behaves like a fluid, and its escape from the system is analogous to the flow of a gas into vacuum from a vessel with a hole. For the system we propose we envisage radially injecting atomic or U238 plasma beam at a pre-determined position and axially pulsing an antiproton beam which upon interaction with the uranium target gives rise to near isotropic ejection of fission fragments with a total mass of 212 amu and total energy of about 160 MeV. These particles, along with the annihilation products (i.e. pions and muons) will heat the background U238 gas - inserted into the chamber just prior to the release of the antiproton - to one keV temperature. Preliminary analysis reveals that such a propulsion system can produce a specific impulse of about 3000 seconds at a thrust of about 50 kN. When applied to a round trip Mars mission, we find that such a journey can be accomplished in about 142 days with 2 days of thrusting and requiring only one gram of antiprotons to achieve it.

  3. RECOVERY OF ALUMINUM FROM FISSION PRODUCTS

    DOEpatents

    Blanco, R.E.; Higgins, I.R.

    1962-11-20

    A method is given for recovertng aluminum values from aqueous solutions containing said values together with fission products. A mixture of Fe/sub 2/O/ sub 3/ and MnO/sub 2/ is added to a solution containing aluminum and fission products. The resulting aluminum-containing supernatant is then separated from the fission product-bearing metal oxide precipitate and is contacted with a cation exchange resin. The aluminum sorbed on the resin is then eluted and recovered. (AEC)

  4. FISSION PRODUCT REMOVAL FROM ORGANIC SOLUTIONS

    DOEpatents

    Moore, R.H.

    1960-05-10

    The decontamination of organic solvents from fission products and in particular the treatment of solvents that were used for the extraction of uranium and/or plutonium from aqueous acid solutions of neutron-irradiated uranium are treated. The process broadly comprises heating manganese carbonate in air to a temperature of between 300 and 500 deg C whereby manganese dioxide is formed; mixing the manganese dioxide with the fission product-containing organic solvent to be treated whereby the fission products are precipitated on the manganese dioxide; and separating the fission product-containing manganese dioxide from the solvent.

  5. METHOD FOR SEPARATING PLUTONIUM AND FISSION PRODUCTS EMPLOYING AN OXIDE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Davies, T.H.

    1961-07-18

    Carrier precipitation processes for separating plutonium values from uranium fission products are described. Silicon dioxide or titanium dioxide in a finely divided state is added to an acidic aqueous solution containing hexavalent plutonium ions together with ions of uranium fission products. The supernatant solution containing plutonium ions is then separated from the oxide and the fission products associated therewith.

  6. Fission Product Sorptivity in Graphite

    SciTech Connect

    Tompson, Jr., Robert V.; Loyalka, Sudarshan; Ghosh, Tushar; Viswanath, Dabir; Walton, Kyle; Haffner, Robert

    2015-04-01

    Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one

  7. PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

    DOEpatents

    Spedding, F.H.; Butler, T.A.; Johns, I.B.

    1959-03-10

    The removal of fission products such as strontium, barium, cesium, rubidium, or iodine from neutronirradiated uranium is described. Uranium halide or elemental halogen is added to melted irradiated uranium to convert the fission products to either more volatile compositions which vaporize from the melt or to higher melting point compositions which separate as solids.

  8. SOURCE OF PRODUCTS OF NUCLEAR FISSION

    DOEpatents

    Harteck, P.; Dondes, S.

    1960-03-15

    A source of fission product recoil energy suitable for use in radiation chemistry is reported. The source consists of thermal neutron irradiated glass wool having a diameter of 1 to 5 microns and containing an isotope fissionable by thermal neutrons, such as U/sup 235/.

  9. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    NASA Technical Reports Server (NTRS)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  10. 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.

  11. Transport properties of fission product vapors

    SciTech Connect

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors.

  12. Modeling Fission Product Sorption in Graphite Structures

    SciTech Connect

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-04-08

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission

  13. 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.

  14. Recovery and use of fission product noble metals

    SciTech Connect

    Jensen, G.A.; Rohmann, C.A.; Perrigo, L.D.

    1980-06-01

    Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value. (DLC)

  15. Fission product release from irradiated LWR fuel under accident conditions

    SciTech Connect

    Strain, R.V.; Sanecki, J.E.; Osborne, M.F.

    1984-01-01

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 2000/sup 0/C are presented in this paper.

  16. Fission Barriers and Neutron Gas in Compound Superheavy Nuclei

    SciTech Connect

    Pei, Junchen; Nazarewicz, W.; Sheikh, J. A.; Kerman, A. K.

    2010-01-01

    Fission and neutron emission are the principal cooling mechanisms of the compound superheavy nuclei. In the framework of the Finite-Temperature Hartree-Fock-Bogoliubov method, the fission barriers and neutron gas have been studied in the excited superheavy systems. Very different energy dependence of fission barriers has been found for ^{278}112 and ^{292}114. On the other hand, the energy dependence of thermal neutron gas has been found to be almost identical for both systems.

  17. A fission gas release correlation for uranium nitride fuel pins

    NASA Technical Reports Server (NTRS)

    Weinstein, M. B.; Davison, H. W.

    1973-01-01

    A model was developed to predict fission gas releases from UN fuel pins clad with various materials. The model was correlated with total release data obtained by different experimentors, over a range of fuel temperatures primarily between 1250 and 1660 K, and fuel burnups up to 4.6 percent. In the model, fission gas is transported by diffusion mechanisms to the grain boundaries where the volume grows and eventually interconnects with the outside surface of the fuel. The within grain diffusion coefficients are found from fission gas release rate data obtained using a sweep gas facility.

  18. Fission gas release restrictor for breached fuel rod

    DOEpatents

    Kadambi, N. Prasad; Tilbrook, Roger W.; Spencer, Daniel R.; Schwallie, Ambrose L.

    1986-01-01

    In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

  19. Fission Product Decay Heat Calculations for Neutron Fission of 232Th

    NASA Astrophysics Data System (ADS)

    Son, P. N.; Hai, N. X.

    2016-06-01

    Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.

  20. Fission product release from TRIGA-LEU reactor fuels

    SciTech Connect

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S

    1980-07-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  1. Rapid separation of fresh fission products (draft)

    SciTech Connect

    Dry, D. E.; Bauer, E.; Petersen, L. 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.

  2. Sensitivity analysis of the fission gas behavior model in BISON.

    SciTech Connect

    Swiler, Laura Painton; Pastore, Giovanni; Perez, Danielle; Williamson, Richard

    2013-05-01

    This report summarizes the result of a NEAMS project focused on sensitivity analysis of a new model for the fission gas behavior (release and swelling) in the BISON fuel performance code of Idaho National Laboratory. Using the new model in BISON, the sensitivity of the calculated fission gas release and swelling to the involved parameters and the associated uncertainties is investigated. The study results in a quantitative assessment of the role of intrinsic uncertainties in the analysis of fission gas behavior in nuclear fuel.

  3. REMOVAL OF FISSION PRODUCTS FROM WATER

    DOEpatents

    Rosinski, J.

    1961-12-19

    A process is given for precipitating fission products from a body of water having a pH of above 6.5. Calcium permanganate and ferrous sulfate are added in a molar ratio of l: 3, whereby a mixed precipitate of manganese dioxide, ferric hydroxide and calcium sulfate is formed; the precipitate carries the fisston products and settles to the bottom of the body of water. (AEC)

  4. Fission-gas release from uranium nitride at high fission rate density

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

    A sweep gas facility has been used to measure the release rates of radioactive fission gases from small UN specimens irradiated to 8-percent burnup at high fission-rate densities. The measured release rates have been correlated with an equation whose terms correspond to direct recoil release, fission-enhanced diffusion, and atomic diffusion (a function of temperature). Release rates were found to increase linearly with burnups between 1.5 and 8 percent. Pore migration was observed after operation at 1550 K to over 6 percent burnup.

  5. Recoil release of fission products from nuclear fuel

    NASA Astrophysics Data System (ADS)

    Wise, C.

    1985-10-01

    An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2. The calculations presented here are one way of allowing for this, other methods are suggested.

  6. Modeling of Fission Gas Release in UO2

    SciTech Connect

    MH Krohn

    2006-01-23

    A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

  7. Mechanistic prediction of fission product release under normal and accident conditions: key uncertainties that need better resolution

    SciTech Connect

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.

  8. Mechanistic prediction of fission-product release under normal and accident conditions: key uncertainties that need better resolution. [PWR; BWR

    SciTech Connect

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.

  9. Measurement of fission product gases in the atmosphere

    NASA Astrophysics Data System (ADS)

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

    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 means of detection, especially for fission gases that decay by beta or electron emission. It is, therefore, customary to concentrate the major gaseous fission products (krypton, xenon and iodine) by cryogenic adsorption for subsequent separation and measurement. This study summarizes our initial efforts to develop an automated portable system for on-line separation and concentration with the potential for measuring environmental levels of radioactive gases, including 85Kr, 131,133,135Xe, 14C, 3H, 35S, 125,131I, etc., without using cryogenic fluids. Bench top and prototype models were constructed using the principle of heatless fractionation of the gases in a pressure swing system. This method removes the requirement for cryogenic fluids to concentrate gases and, with suitable electron and gamma ray detectors, provides for remote use under automatic computer control. Early results using 133Xe tracer show that kinetic chromatography, i.e., high pressure adsorption of xenon and low pressure desorption of air, using specific types of molecular sieves, permits the separation and quantification of xenon isotopes from large volume air samples. We are now developing the ability to measure the presence and amounts of fission-produced xenon isotopes that decay by internal conversion electrons and beta radiation with short half-lives, namely 131mXe, 11.8 d, 133mXe, 2.2 d, 133Xe, 5.2 d and 135Xe, 9.1 h. The ratio of the isotopic concentrations measured can be used to determine unequivocally the amount of fission gas and time of release of an air parcel many kilometers downwind from a

  10. DESIGN OF AN ON-LINE, MULTI-SPECTROMETER FISSION PRODUCT MONITORING SYSTEM (FPMS) TO SUPPORT ADVANCED GAS REACTOR (AGR) FUEL TESTING AND QUALIFICATION IN THE ADVANCED TEST REACTOR

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert

    2005-11-01

    The US Department of Energy (DOE) is embarking on a series of tests of coated-particle reactor fuel for the Advanced Gas Reactor (AGR). As one part of this fuel development program a series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). The first test in this series (AGR-1) will incorporate six separate “capsules” irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitored in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed, and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the AGR-1 fission product measurement system (FPMS) have sufficient sensitivity to detect the failure of a single coated fuel particle and sufficient range to allow it to “count” multiple (up to 250) successive particle failures. This paper describes the design and expected performance of the AGR-1 FPMS.

  11. Calculations on fission gas behaviour in the high burnup structure

    NASA Astrophysics Data System (ADS)

    Blair, P.; Romano, A.; Hellwig, Ch.; Chawla, R.

    2006-05-01

    The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO 2 fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.

  12. 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.

  13. ORNL fission product release tests VI-6

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Collins, J.L.; Lee, C.S.

    1991-01-01

    The ORNL fission product release tests investigate release and transport of the major fission products from high-burnup fuel under LWR accident conditions. The two most recent tests (VI-4 and VI-5) were conducted in hydrogen. In three previous tests in this series (VI-1, VI-2, and VI-3), which had been conducted in steam, the oxidized Zircaloy cladding remained largely intact and acted as a barrier to steam reaction with the UO{sub 2}. Test VI-6 was designed to insure significant oxidation of the UO{sub 2} fuel, which has been shown to enhance release of certain fission products, especially molybdenum and ruthenium. The BR3 fuel specimen used in test VI-6 will be heated in hydrogen to 2300 K; the Zircaloy cladding is expected to melt and runoff at {approximately}2150 K. Upon reaching the 2300 K test temperature, the test atmosphere will be changed to steam, and that temperature will be maintained for 60 min, with the three collection trains being operated for 2-, 18-, and 40-min periods. The releases of {sup 85}Kr and {sup 137}Cs will be monitored continuously throughout the test. Posttest analyses of the material collected on the three trains will provide results on the release and transport of Mo, Ru, Sb, Te, Ba, Ce, and Eu as a function of time at 2300 K. Continuous monitoring of the hydrogen produced during the steam atmosphere period at high temperature will provide a measure of the oxidation rate of the cladding and fuel. Following delays in approval of the safety documentation and in decontamination of the hot cell and test apparatus, test VI-6 will be conducted in late May.

  14. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

    DOEpatents

    Nicholls, C.M.; Wells, I.; Spence, R.

    1959-10-13

    The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

  15. Decay Chain Deduction of Uranium Fission Products.

    PubMed

    Guo, Huiping; Tian, Chenyang; Wang, Xiaotian; Lv, Ning; Ma, Meng; Wei, Yingguang

    2016-07-01

    Delayed gamma spectrum is the fingerprint of uranium materials in arms control verification technology. The decay chain is simplified into basic state linear chain and excitation state linear chain to calculate and analyze the delayed gamma spectra of fission products. Formulas of the changing rule for nuclide number before and after zero-time are deduced. The C program for calculating the delayed gamma ray spectra data is constructed, and related experiments are conducted to verify this theory. Through analysis of the delayed gamma counts of several nuclides, the calculated results are found to be consistent with experimental values. PMID:27218290

  16. Data summary report for fission product release test VI-6

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L.

    1994-03-01

    Test VI-6 was the sixth 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 BR3 reactor in Belgium. The fuel had experienced a burnup of {approximately}42 MWd/kg, with inert gas release during irradiation of {approximately}2%. The fuel specimen was heated in an induction furnace at 2300 K for 60 min, initially in hydrogen, then in a steam atmosphere. The released fission products were collected in three sequentially operated collection trains designed to facilitate sampling and analysis. The fission product inventories in the fuel were measured directly by gamma-ray spectrometry, where possible, and were calculated by ORIGEN2. Integral releases were 75% for {sup 85}Kr, 67% for {sup 129}I, 64% for {sup 125}Sb, 80% for both {sup 134}Cs and {sup 137}Cs, 14% for {sup 154}Eu, 63% for Te, 32% for Ba, 13% for Mo, and 5.8% for Sr. Of the totals released from the fuel, 43% of the Cs, 32% of the Sb, and 98% of the Eu were deposited in the outlet end of the furnace. During the heatup in hydrogen, the Zircaloy cladding melted, ran down, and reacted with some of the UO{sub 2} and fission products, especially Te and Sb. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.57 g, almost equally divided between thermal gradient tubes and filters. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL Diffusion Model.

  17. Energy production using fission fragment rockets

    NASA Astrophysics Data System (ADS)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

  18. Ceramic Hosts for Fission Products Immobilization

    SciTech Connect

    Peter C Kong

    2010-07-01

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  19. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

    DOEpatents

    Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

    1958-10-01

    A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

  20. Fission product plateout/liftoff/washoff test plan. Revision 1

    SciTech Connect

    Acharya, R.; Hanson, D.

    1988-05-01

    A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.

  1. Transmission electron microscopy characterization of the fission gas bubble superlattice in irradiated U-7wt% Mo dispersion fuels

    SciTech Connect

    B.D. Miller; J. Gan; D.D. Keiser Jr.; A.B. Robinson; J.-F. Jue; J.W. Madden; P.G. Medvedev

    2015-03-01

    Transmission electron microscopy characterization of irradiated U-7wt% Mo dispersion fuel was performed on various samples to understand the effect of irradiation parameters (fission density, fission rate, and temperature) on the self-organized fission-gas-bubble superlattice that forms in the irradiated U-Mo fuel. The bubble superlattice was seen to form a face-centered cubic structure coherent with the host U-7wt% Mo body centered cubic structure. At a fission density between 3.0 and 4.5 x 1021 fiss/cm3, the superlattice bubbles appear to have reached a saturation size with additional fission gas associated with increasing burnup predominately accumulating along grain boundaries. At a fission density of ~4.5x1021 fiss/cm3, the U-7wt% Mo microstructure undergoes grain subdivision and can no longer support the ordered bubble superlattice. The fuel grains are primarily less than 500 nm in diameter with micron-size fission-gas bubbles present on the grain boundaries. Solid fission products decorate the inside surface of the micron-sized fission-gas bubbles. Residual superlattice bubbles are seen in areas where fuel grains remain micron sized. Potential mechanisms of the formation and collapse of the bubble superlattice are discussed.

  2. Consideration of Grain Size Distribution in the Diffusion of Fission Gas to Grain Boundaries

    SciTech Connect

    Paul C. Millett; Yongfeng Zhang; Michael R. Tonks; S. B. Biner

    2013-09-01

    We analyze the accumulation of fission gas on grain boundaries in a polycrystalline microstructure with a distribution of grain sizes. The diffusion equation is solved throughout the microstructure to evolve the gas concentration in space and time. Grain boundaries are treated as infinite sinks for the gas concentration, and we monitor the cumulative gas inventory on each grain boundary throughout time. We consider two important cases: first, a uniform initial distribution of gas concentration without gas production (correlating with post-irradiation annealing), and second, a constant gas production rate with no initial gas concentration (correlating with in-reactor conditions). The results show that a single-grain-size model, such as the Booth model, over predicts the gas accumulation on grain boundaries compared with a polycrystal with a grain size distribution. Also, a considerable degree of scatter, or variability, exists in the grain boundary gas accumulation when comparing all of the grain boundaries in the microstructure.

  3. Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

    2006-10-01

    The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

  4. (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.

  5. Fission products stability in uranium dioxide

    NASA Astrophysics Data System (ADS)

    Brillant, G.; Gupta, F.; Pasturel, A.

    2011-05-01

    Fission product stability in nuclear fuels is investigated using density functional theory (DFT). In particular, incorporation and solution energies of He, Kr, Xe, I, Te, Ru, Sr and Ce in pre-existing trap sites of UO 2 (vacancies, interstitials, U-O divacancy, and Schottky trio defects) are calculated using the projector-augmented-wave method as implemented in the Vienna ab initio simulation package. Correlation effects are taken into account within the DFT+U approach. The stability of many binary and ternary compounds in comparison to soluted atoms is also explored. Finally the involvement of FP in the formation of metallic and oxide precipitates in oxide fuels is discussed in the light of experimental results.

  6. REGENERATION OF FISSION-PRODUCT-CONTAINING MAGNESIUM-THORIUM ALLOYS

    DOEpatents

    Chiotti, P.

    1964-02-01

    A process of regenerating a magnesium-thorium alloy contaminated with fission products, protactinium, and uranium is presented. A molten mixture of KCl--LiCl-MgCl/sub 2/ is added to the molten alloy whereby the alkali, alkaline parth, and rare earth fission products (including yttrium) and some of the thorium and uranium are chlorinated and

  7. SEPARATION OF FISSION PRODUCTS FROM PLUTONIUM BY PRECIPITATION

    DOEpatents

    Seaborg, G.T.; Thompson, S.G.; Davidson, N.R.

    1959-09-01

    Fission product separation from hexavalent plutonium by bismuth phosphate precipitation of the fission products is described. The precipitation, according to this invention, is improved by coprecipitating ceric and zirconium phosphates (0.05 to 2.5 grams/liter) with the bismuth phosphate.

  8. Immobilization of fission products in phosphate ceramic waste forms

    SciTech Connect

    Singh, D.

    1996-10-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products.

  9. Bulk and surface controlled diffusion of fission gas atoms

    SciTech Connect

    Andersson, Anders D.

    2012-08-09

    Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO

  10. (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.

  11. Fission-product SiC reaction in HTGR fuel

    SciTech Connect

    Montgomery, F.

    1981-07-13

    The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels.

  12. 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.

  13. Fission Product Transmutation in Mixed Radiation Fields

    SciTech Connect

    Harmon, Frank; Burgett, Erick; Starovoitova, Valeriia; Tsveretkov, Pavel

    2015-01-15

    Work under this grant addressed a part of the challenge facing the closure of the nuclear fuel cycle; reducing the radiotoxicity of lived fission products (LLFP). It was based on the possibility that partitioning of isotopes and accelerator-based transmutation on particular LLFP combined with geological disposal may lead to an acceptable societal solution to the problem of management. The feasibility of using photonuclear processes based on the excitation of the giant dipole resonance (GDR) by bremsstrahlung radiation as a cost effective transmutation method was accessed. The nuclear reactions of interest: (γ,xn), (n,γ), (γ,p) can be induced by bremsstrahlung radiation produced by high power electron accelerators. The driver of these processes would be an accelerator that produces a high energy and high power electron beam of ~ 100 MeV. The major advantages of such accelerators for this purpose are that they are essentially available “off the shelf” and potentially would be of reasonable cost for this application. Methods were examined that used photo produced neutrons or the bremsstrahlung photons only, or use both photons and neutrons in combination for irradiations of selected LLFP. Extrapolating the results to plausible engineering scale transmuters it was found that the energy cost for 129I and 99Tc transmutation by these methods are about 2 and 4%, respectively, of the energy produced from 1000MWe.

  14. Fission gas bubble percolation on crystallographically consistent grain boundary networks

    NASA Astrophysics Data System (ADS)

    Sabogal-Suárez, Daniel; David Alzate-Cardona, Juan; Restrepo-Parra, Elisabeth

    2016-07-01

    Fission gas release in nuclear fuels can be modeled in the framework of percolation theory, where each grain boundary is classified as open or closed to the release of the fission gas. In the present work, two-dimensional grain boundary networks were assembled both at random and in a crystallographically consistent manner resembling a general textured microstructure. In the crystallographically consistent networks, grain boundaries were classified according to its misorientation. The percolation behavior of the grain boundary networks was evaluated as a function of radial cracks and radial thermal gradients in the fuel pellet. Percolation thresholds tend to shift to the left with increasing length and number of cracks, especially in the presence of thermal gradients. In general, the topology and percolation behavior of the crystallographically consistent networks differs from those of the random network.

  15. Diffusion of Zr, Ru, Ce, Y, La, Sr and Ba fission products in UO2

    NASA Astrophysics Data System (ADS)

    Perriot, R.; Liu, X.-Y.; Stanek, C. R.; Andersson, D. A.

    2015-04-01

    The diffusivity of the solid fission products (FP) Zr (Zr4+), Ru (Ru4+, Ru3+), Ce (Ce4+), Y (Y3+), La (La3+), Sr (Sr2+) and Ba (Ba2+) by a vacancy mechanism has been calculated, using a combination of density functional theory (DFT) and empirical potential (EP) calculations. The activation energies for the solid fission products are compared to the activation energy for Xe fission gas atoms calculated previously. Apart from Ru, the solid fission products all exhibit higher activation energy than Xe. For all solid FPs except Y3+, the migration of the FP has lower barrier than the migration of a neighboring U atom, making the latter the rate limiting step for direct migration. An indirect mechanism, consisting of two successive migrations around the FP, is also investigated. The calculated diffusivities show that most solid fission products diffuse with rates similar to U self-diffusion. However, Ru, Ba and Sr exhibit faster diffusion than the other solid FPs, with Ru3+ and Ru4+ diffusing even faster than Xe for T < 1200 K. The diffusivities correlate with the observed fission product solubility in UO2, and the tendency to form metallic and oxide second phase inclusions.

  16. Fission gas bubble identification using MATLAB's image processing toolbox

    DOE PAGESBeta

    Collette, R.; King, J.; Keiser, Jr., D.; Miller, B.; Madden, J.; Schulthess, J.

    2016-06-08

    Automated image processing routines have the potential to aid in the fuel performance evaluation process by eliminating bias in human judgment that may vary from person-to-person or sample-to-sample. In addition, this study presents several MATLAB based image analysis routines designed for fission gas void identification in post-irradiation examination of uranium molybdenum (U–Mo) monolithic-type plate fuels. Frequency domain filtration, enlisted as a pre-processing technique, can eliminate artifacts from the image without compromising the critical features of interest. This process is coupled with a bilateral filter, an edge-preserving noise removal technique aimed at preparing the image for optimal segmentation. Adaptive thresholding provedmore » to be the most consistent gray-level feature segmentation technique for U–Mo fuel microstructures. The Sauvola adaptive threshold technique segments the image based on histogram weighting factors in stable contrast regions and local statistics in variable contrast regions. Once all processing is complete, the algorithm outputs the total fission gas void count, the mean void size, and the average porosity. The final results demonstrate an ability to extract fission gas void morphological data faster, more consistently, and at least as accurately as manual segmentation methods.« less

  17. Thermodynamics of fission products in UO2+-x

    SciTech Connect

    Nerikar, Pankaj V

    2009-01-01

    The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

  18. TRAMP. Transport of Metallic Fission Products Along Multiple Parallel Paths

    SciTech Connect

    Hudritsch, W.; Richards, M.

    1991-11-01

    TRAMP is used to calculate the transport of metallic fission products along multiple parallel paths; the primary application is transport in and release from nuclear-grade graphite. The transport mechanisms are concentration-driven diffusion, thermal diffusion, and convection.

  19. 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 INL’s 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.

  20. Fission product studies in the symmetric mass region

    SciTech Connect

    De Laeter, J.R.; Rosman, K.J.R.; Loss, R.D.

    1993-05-01

    Fission yields can be determined by radiochemical or mass spectrometric techniques. Mass spectrometry can provide more accurate data, particularly in the symmetric mass region where the probability of fission is low and uncertainties in isometric ratios occur. Fine structure in the mass distribution can usually only be determined by mass spectrometry. Many of the elements in the valley of symmetry have high ionization potentials and are therefore difficult to measure by solid source mass spectrometry. Analytical techniques have been developed to provide the sensitivity required to measure the small sample sizes available in fission product studies. Cumulative fission yields for ruthenium, palladium, cadmium, tin, and tellurium have been measured by mass spectrometry for the thermal and epicadmium fission of {sup 233}U and for thermal and epicadmium fission of {sup 239}Pu. These fission yields, which span the mass range 101 {le} A {le} 130, can be combined to give a mass yield curve for {sup 235}U in the valley region, which is symmetrical about A = 116.8 and exhibits fine structure in the mass 113 to 114 region. Fine structure in {sup 233}U is also present at mass 111. Mass spectrometric determinations of the fission yields of uranium ore at the Oklo mine site in Gabon enable the nuclear parameters of this natural reactor to be evaluated. This in turn enables the amounts of fission products produced in the reactor zone and the surrounding rocks enables an assessment to be made of the efficiency of this geological repository for containing radioactive waste. The elemental abundances can be determined by isotope dilution mass spectrometry. Unfortunately, the paucity of good fission yield data available for {sup 238}U by fast neutrons is a severe constraint in this evaluation.

  1. Development of a multiscale thermal conductivity model for fission gas in UO2

    NASA Astrophysics Data System (ADS)

    Tonks, Michael R.; Liu, Xiang-Yang; Andersson, David; Perez, Danielle; Chernatynskiy, Aleksandr; Pastore, Giovanni; Stanek, Christopher R.; Williamson, Richard

    2016-02-01

    Accurately predicting changes in the thermal conductivity of light water reactor UO2 fuel throughout its lifetime in reactor is an essential part of fuel performance modeling. However, typical thermal conductivity models from the literature are empirical. In this work, we begin to develop a mechanistic thermal conductivity model by focusing on the impact of gaseous fission products, which is coupled to swelling and fission gas release. The impact of additional defects and fission products will be added in future work. The model is developed using a combination of atomistic and mesoscale simulation, as well as analytical models. The impact of dispersed fission gas atoms is quantified using molecular dynamics simulations corrected to account for phonon-spin scattering. The impact of intragranular bubbles is accounted for using an analytical model that considers phonon scattering. The impact of grain boundary bubbles is determined using a simple model with five thermal resistors that are parameterized by comparing to 3D mesoscale heat conduction results. When used in the BISON fuel performance code to model four reactor experiments, it produces reasonable predictions without having been fit to fuel thermocouple data.

  2. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  3. 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 1.2×10 2 (Capsule 3) to 3.8×10 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.

  4. 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.

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

    SciTech Connect

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  6. Evaluation and compilation of fission product yields 1993

    SciTech Connect

    England, T.R.; Rider, B.F.

    1995-12-31

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

  7. 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.

  8. 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.

  9. An efficient model for the analysis of fission gas release

    NASA Astrophysics Data System (ADS)

    Bernard, L. C.; Jacoud, J. L.; Vesco, P.

    2002-04-01

    This paper presents the fission gas release (FGR) model that has been developed at Framatome ANP and incorporated into its fuel rod performance code COPERNIC in order to accurately predict FGR into pressurized water reactor fuel rods under normal and off-normal operating conditions including UO 2, gadolinia and MOX fuels. The model is analytical, thus enabling fast and robust fuel rod calculations, a must within an industrial framework where safety evaluations may require the analyses of a full core and of a very large number of transients. Although the model is simple, it includes the most important FGR features: athermal, thermal, steady-state, and transient regimes, burst effect, rim formation, and MOX-type microstructure. The validation of the model covers 400 irradiated rods that include high burnups, high powers, short to long transients, and shows the quality of the prediction of the model in all types of conditions. As temperature is a key parameter that affects FGR, the COPERNIC thermal model is briefly described and its impact on fission gas released uncertainty is discussed.

  10. 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

  11. FASTGRASS implementation in BISON and Fission gas behavior characterization in UO2 and connection to validating MARMOT

    SciTech Connect

    Yun, Di; Mo, Kun; Ye, Bei; Jamison, Laura M.; Miao, Yinbin; Lian, Jie; Yao, Tiankei

    2015-09-30

    This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL). Two major accomplishments in FY 15 are summarized in this report: (1) implementation of the FASTGRASS module in the BISON code; and (2) a Xe implantation experiment for large-grained UO2. Both BISON AND MARMOT codes have been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. To contribute to the development of the Moose-Bison-Marmot (MBM) code suite, we have implemented the FASTGRASS fission gas model as a module in the BISON code. Based on rate theory formulations, the coupled FASTGRASS module in BISON is capable of modeling LWR oxide fuel fission gas behavior and fission gas release. In addition, we conducted a Xe implantation experiment at the Argonne Tandem Linac Accelerator System (ATLAS) in order to produce the needed UO2 samples with desired bubble morphology. With these samples, further experiments to study the fission gas diffusivity are planned to provide validation data for the Fission Gas Release Model in MARMOT codes.

  12. Fission product release from nuclear fuel by recoil and knockout

    NASA Astrophysics Data System (ADS)

    Lewis, B. J.

    1987-03-01

    An analytical model has been developed to describe the fission product release from nuclear fuel arising from the surface-fission release mechanisms of recoil and knockout. Release expressions are evaluated and compared to the short-lived activity measurements from in-reactor experiments with intact operating fuel. Recoil is shown to be an important process for releasing fission products from free UO 2 surfaces into the fuel-to-sheath gap. The model is also applied to tramp uranium in a power reactor primary heat transport circuit where it is demonstrated that recoil is the dominant release mechanism for small particles of fuel which are deposited on in-core surfaces. A methodology is established whereby release from surface contamination can be distinguished from that of fuel pin failure.

  13. (Fission product transport experiments (HFR-B1))

    SciTech Connect

    Myers, B.F.

    1989-12-05

    Travel to the JRC Petten was for the purpose of discussing the HFR-B1 experiment and post irradiation activities. Technical assessment of the experiment strongly supports the concept of enhanced fission gas release at temperatures above 1100{degree}C, the extensive release of stored fission gas at water vapor levels postulated in accident scenarios, an increase in the steady-state fission gas release under hydrolyzing conditions, and an increase in gas release during thermal cycling. Schedules were established for completion of the work and issuance of reports by September 1990. At the KFA Juelich agreement was reached on the PIE activities for HFR-B1 and a schedule established. The final PIE report is due June 1991. Choices of accident condition tests in the PIE have yet to be made by the US participants. A proposal for the establishment of a new cooperative effort on model and code development was presented at the Institut fuer Nukleare Sicherheitsforschung of KFA. The proposal was considered premature; discussions dealing with general principles, basic aims, and organization were requested; particular concerns about free exchange of information, overlap with the existing safety subprogram, and exclusive cooperation with ORNL were raised. A strong desire for cooperation and the opinion that the raised problems could be resolved were expressed. Technical discussions at the KFA were beneficial.

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

    SciTech Connect

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

    2011-07-01

    a gamma spectrometer in the fission product laboratory of the future Jules Horowitz Material Test Reactor. An intermediate step will consist of testing first equipment on an existing experimental facility in the LECA-STAR Hot Cell Laboratory of the CEA Cadarache. This paper presents the scientific and operational stakes linked to fission gas issues, resumes the current state of art for analyzing them in nuclear facilities, then presents the skills gathered through this collaboration to overcome technological bottlenecks. Finally it describes the implementation strategy in nuclear research facilities of the CEA Cadarache. (authors)

  15. Deposition of volatile fission products during commercial high-level waste vitrification

    SciTech Connect

    Hanson, M.S.; Carter, J.G.

    1982-09-01

    High-level waste from spent commercial PWR fuel was vitrified during the Nuclear Waste Vitrification Project (NWVP), conducted at Pacific Northwest Laboratory in 1978 and 1979. The solidification equipment employed sintered metal filters and a conventional wet off-gas cleanup system to decontaminate the gases generated during waste processing. This report is a summary of a study completed under contract with U.S. Department of Energy (DOE) to determine the fate of volatile fission products in the filters and off-gas system. The study objectives were to determine (1) the extent of volatile fission product deposition within the sintered metal filters, (2) the distribution of fission products in the off-gas system, and (3) the factors that control the distribution. Such information is needed for the design and operation of future waste immobilization systems. The scope of the study included identification, preparation, and microprobe analyses of suitable filter samples and radiochemical analyses of off-gas system samples. The report describes the methods of sample selection and gathering, the analytical methods and results, and the interpretations of the results. Relationships to vitrification system operations are defined and operating conditions to minimize problems recommended. 37 figures, 14 tables.

  16. Data summary report for fission product release test VI-5

    SciTech Connect

    Osborne, M.F.; Lorenz, R.A.; Travis, J.R.; Webster, C.S.; Collins, J.L. )

    1991-10-01

    Test VI-5, the fifth in a series of high-temperature fission product release tests in a vertical test apparatus, was conducted in a flowing mixture of hydrogen and helium. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium which had been irradiated to a burnup of {approximately}42 MWd/kg. Using a hot cell-mounted test apparatus, the fuel rod was heated in an induction furnace under simulated LWR accident conditions to two test temperatures, 2000 K for 20 min and then 2700 K for an additional 20 min. The released fission products were collected in three sequentially operated collection trains on components designed to measure fission product transport characteristics and facilitate sampling and analysis. The results from this test were compared with those obtained in previous tests in this series and with the CORSOR-M and ORNL diffusion release models for fission product release. 21 refs., 19 figs., 12 tabs.

  17. Yields of Fission Products from Various Uranium and Thorium Targets

    SciTech Connect

    Kronenberg, Andreas; Spejewski, Eugene H.; Mervin, Brenden T.; Jost, Cara; Carter, H Kennon; Stracener, Daniel W; Greene, John P.; Nolen, Jerry A.; Talbert, Willard L.

    2008-01-01

    Yield measurements from proton-induced fission have been performed on a number of actinide targets, both Th and U, at the on-line test facility at Oak Ridge National Laboratory. The results are discussed with a focus on the production process and physical and chemical properties of the targets.

  18. Yields of fission products from various uranium and thorium targets.

    SciTech Connect

    Kronenberg, A.; Spejewski, E. H.; Mervin, B.; Jost, C.; Carter, H. K.; Stracener, D. W.; Greene, J. P.; Nolen, J. A.; Talbert, W. L.; Physics; Oak Ridge Associated Univ.; ORNL; TechSource, Inc.

    2008-10-31

    Yield measurements from proton-induced fission have been performed on a number of actinide targets, both Th and U, at the on-line test facility at Oak Ridge National Laboratory. The results are discussed with a focus on the production process and physical and chemical properties of the targets.

  19. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

    DOEpatents

    Seaborg, G.T.; Willard, J.E.

    1958-01-01

    A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

  20. 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)

  1. Applications for fission product data to problems in stellar nucleosynthesis

    SciTech Connect

    Mathews, G.J.

    1983-10-01

    A general overview of the nucleosynthesis mechanisms for heavy (A greater than or equal to 70) nuclei is presented with particular emphasis on critical data needs. The current state of the art in nucleosynthesis models is described and areas in which fission product data may provide useful insight are proposed. 33 references, 10 figures.

  2. A Covariance Generation Methodology for Fission Product Yields

    NASA Astrophysics Data System (ADS)

    Terranova, N.; Serot, O.; Archier, P.; Vallet, V.; De Saint Jean, C.; Sumini, M.

    2016-03-01

    Recent safety and economical concerns for modern nuclear reactor applications have fed an outstanding interest in basic nuclear data evaluation improvement and completion. It has been immediately clear that the accuracy of our predictive simulation models was strongly affected by our knowledge on input data. Therefore strong efforts have been made to improve nuclear data and to generate complete and reliable uncertainty information able to yield proper uncertainty propagation on integral reactor parameters. Since in modern nuclear data banks (such as JEFF-3.1.1 and ENDF/BVII.1) no correlations for fission yields are given, in the present work we propose a covariance generation methodology for fission product yields. The main goal is to reproduce the existing European library and to add covariance information to allow proper uncertainty propagation in depletion and decay heat calculations. To do so, we adopted the Generalized Least Square Method (GLSM) implemented in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation), developed at CEA-Cadarache. Theoretical values employed in the Bayesian parameter adjustment are delivered thanks to a convolution of different models, representing several quantities in fission yield calculations: the Brosa fission modes for pre-neutron mass distribution, a simplified Gaussian model for prompt neutron emission probability, theWahl systematics for charge distribution and the Madland-England model for the isomeric ratio. Some results will be presented for the thermal fission of U-235, Pu-239 and Pu-241.

  3. Recent MELCOR and VICTORIA Fission Product Research at the NRC

    SciTech Connect

    Bixler, N.E.; Cole, R.K.; Gauntt, R.O.; Schaperow, J.H.; Young, M.F.

    1999-01-21

    The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes in the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02

  4. 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.

  5. Fission Product Yields from Fission Spectrum n+ 239Pu for ENDF/B-VII.1

    NASA Astrophysics Data System (ADS)

    Chadwick, M. B.; Kawano, T.; Barr, D. W.; Mac Innes, M. R.; Kahler, A. C.; Graves, T.; Selby, H.; Burns, C. J.; Inkret, W. C.; Keksis, A. L.; Lestone, J. P.; Sierk, A. J.; Talou, P.

    2010-12-01

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for 99Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on

  6. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  7. Design of pellet surface grooves for fission gas plenum

    SciTech Connect

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM.

  8. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  9. 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

  10. Characterization of intergranular fission gas bubbles in U-Mo fuel.

    SciTech Connect

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.; Nuclear Engineering Division; SSCR RIAR

    2008-04-14

    plates were also performed and compared with the calculated quantities. Bubble density per unit grain boundary surface area was calculated by using the density per unit grain boundary length data. These data were used as input for mechanistic modeling described in section 4. Volumetric bubble density was calculated by using density per unit grain boundary surface area. Based on these data, bubble volumetric fraction was calculated. Bubble volume fraction was also calculated by using the density per unit cross section area. Bubble volume fraction was also directly measured for some plates. These three results are comparable although the direct measurement data are slightly larger than the others. Bubble volume fraction increased as a function of burnup, reaching {approx}2% of fuel volume at 3 x 10{sup 21} f/cm{sup 3}. Fission gas bubble swelling is minor compared to that of solid fission product swelling.

  11. Measurement of fission gas release from irradiated U-Mo monolithic fuel samples

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  12. Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

    SciTech Connect

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium–molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium–molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  13. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

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

    NASA Astrophysics Data System (ADS)

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

    2015-01-01

    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 with a recently implemented physics-based model for 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 in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions 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, significantly 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.

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

    SciTech Connect

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    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.

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

    DOE PAGESBeta

    Pastore, Giovanni; Swiler, L. P.; Hales, Jason D.; Novascone, Stephen R.; Perez, Danielle M.; Spencer, Benjamin W.; Luzzi, Lelio; Uffelen, Paul Van; Williamson, Richard L.

    2014-10-12

    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 uncertaintymore » 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.« less

  17. Analysis of fission product release behavior during the TMI-2 accident

    SciTech Connect

    Petti, D. A.; Adams, J. P.; Anderson, J. L.; Hobbins, R. R.

    1987-01-01

    An analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been initiated to provide an understanding of fission product behavior that is consistent with both the best estimate accident scenario and fission product results from the ongoing sample acquisition and examination efforts. ''First principles'' fission product release models are used to describe release from intact, disrupted, and molten fuel. Conclusions relating to fission product release, transport, and chemical form are drawn. 35 refs., 12 figs., 7 tabs.

  18. Superabsorbing gel for actinide, lanthanide, and fission product decontamination

    DOEpatents

    Kaminski, Michael D.; Mertz, Carol J.

    2016-06-07

    The present invention provides an aqueous gel composition for removing actinide ions, lanthanide ions, fission product ions, or a combination thereof from a porous surface contaminated therewith. The composition comprises a polymer mixture comprising a gel forming cross-linked polymer and a linear polymer. The linear polymer is present at a concentration that is less than the concentration of the cross-linked polymer. The polymer mixture is at least about 95% hydrated with an aqueous solution comprising about 0.1 to about 3 percent by weight (wt %) of a multi-dentate organic acid chelating agent, and about 0.02 to about 0.6 molar (M) carbonate salt, to form a gel. When applied to a porous surface contaminated with actinide ions, lanthanide ions, and/or other fission product ions, the aqueous gel absorbs contaminating ions from the surface.

  19. Reaction behavior of zircaloy with simulated fission products

    SciTech Connect

    Kohli, R.

    1981-01-01

    The investigation reported in this paper was prompted by the lack of information on the reaction behavior of Zircaloy on long-term exposure to fission product environments in the temperature range 573 - 973 K. Small Zircaloy-2 (Zircaloy-2 contains by weight 1.5% Sn; 0.15% Fe; 0.08% Cr; 0.05% Ni; rest Zr) strip specimens (ca. 25mm x 3mm x 0.75 mm thick) were exposed to various simulated fission product environments (Cs, I, Br, Cd, In, Sb, Sn, Se as vapor; all others as powders) for times to 5.4 Ms (1500 h) in the temperature range 673 - 973 K. The reaction behavior was characterized by scanning electron microscopy with an EDAX analyzer, optical metallography, and x-ray diffraction.

  20. Fission-product release from TRIGA-LEU reactor fuels

    SciTech Connect

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-11-01

    The release of fission products, both gaseous and volatile metals, from TRIGA fuel is important for the analysis of possible accident conditions related to reactor operation and the design of future TRIGA fuel systems. Because of present national concerns over nuclear proliferation, it has become clear that future reactor fuels will, of necessity, utilize low-enriched uranium (LEU, enrichment <20%). This will require increasing the total uranium loading per unit volume of the higher-loaded TRIGA fuels for the purpose of maintaining the appropriate fissile loading. Because of these new developments, tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing 8.5 to 45 wt % uranium.

  1. CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

    SciTech Connect

    Chrien, R.E.; Burrows, T.W.

    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)

  3. Fission track astrology of three Apollo 14 gas-rich breccias

    NASA Technical Reports Server (NTRS)

    Graf, H.; Shirck, J.; Sun, S.; Walker, R.

    1973-01-01

    The three Apollo 14 breccias 14301, 14313, and 14318 all show fission xenon due to the decay of Pu-244. To investigate possible in situ production of the fission gas, an analysis was made of the U-distribution in these three breccias. The major amount of the U lies in glass clasts and in matrix material and no more than 25% occurs in distinct high-U minerals. The U-distribution of each breccia is discussed in detail. Whitlockite grains in breccias 14301 and 14318 found with the U-mapping were etched and analyzed for fission tracks. The excess track densities are much smaller than indicated by the Xe-excess. Because of a preirradiation history documented by very high track densities in feldspar grains, however, it is impossible to attribute the excess tracks to the decay of Pu-244. A modified track method has been developed for measuring average U-concentrations in samples containing a heterogeneous distribution of U in the form of small high-U minerals. The method is briefly discussed, and results for the rocks 14301, 14313, 14318, 68815, 15595, and the soil 64421 are given.

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

    SciTech Connect

    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.

  5. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    SciTech Connect

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors

  6. PolyPole-1: An accurate numerical algorithm for intra-granular fission gas release

    NASA Astrophysics Data System (ADS)

    Pizzocri, D.; Rabiti, C.; Luzzi, L.; Barani, T.; Van Uffelen, P.; Pastore, G.

    2016-09-01

    The transport of fission gas from within the fuel grains to the grain boundaries (intra-granular fission gas release) is a fundamental controlling mechanism of fission gas release and gaseous swelling in nuclear fuel. Hence, accurate numerical solution of the corresponding mathematical problem needs to be included in fission gas behaviour models used in fuel performance codes. Under the assumption of equilibrium between trapping and resolution, the process can be described mathematically by a single diffusion equation for the gas atom concentration in a grain. In this paper, we propose a new numerical algorithm (PolyPole-1) to efficiently solve the fission gas diffusion equation in time-varying conditions. The PolyPole-1 algorithm is based on the analytic modal solution of the diffusion equation for constant conditions, combined with polynomial corrective terms that embody the information on the deviation from constant conditions. The new algorithm is verified by comparing the results to a finite difference solution over a large number of randomly generated operation histories. Furthermore, comparison to state-of-the-art algorithms used in fuel performance codes demonstrates that the accuracy of PolyPole-1 is superior to other algorithms, with similar computational effort. Finally, the concept of PolyPole-1 may be extended to the solution of the general problem of intra-granular fission gas diffusion during non-equilibrium trapping and resolution, which will be the subject of future work.

  7. Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V.

    Energy Science and Technology Software Center (ESTSC)

    1982-10-12

    Version 02 The ENDF/B-IV fission-product files contain neutron cross sections, decay constants, decay energies, and other decay data for 824 important fission products. They also contain fission yields for these fission products produced by one or more fission-neutron energies (14 MeV, fast, and thermal fission). Also, spectral data exist for the most important decay-heat contributors among the 824 nuclides. Because the spectra are based on fission pulses, the libraries have a general utility. The exponentialmore » fits, for example, can be folded into any power (fission) history that can be described analytically or by a histogram representation. The effects of neutron absorption are also treated and approximately accounted for in the methodology.« less

  8. Photofission observations in reactor environments using selected fission-product yields

    SciTech Connect

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1982-01-22

    A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized.

  9. 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.

  10. 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.

  11. 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).

  12. Methods to Collect, Compile, and Analyze Observed Short-lived Fission Product Gamma Data

    SciTech Connect

    Finn, Erin C.; Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.; Ellis, Tere A.

    2011-09-29

    A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

  13. 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.

  14. Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

    SciTech Connect

    Rest, J.; Zawadski, S.A.; Piasecka, M.

    1983-10-01

    The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures.

  15. PURIFICATION OF PLUTONIUM USING A CERIUM PRECIPITATE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Faris, B.F.; Olson, C.M.

    1961-07-01

    Bismuth phosphate carrier precipitation processes are described for the separation of plutonium from fission products wherein in at least one step bismuth phosphate is precipitated in the presence of hexavalent plutonium thereby carrying a portion of the fission products from soluble plu tonium values. In this step, a cerium phosphate precipitate is formed in conjunction with the bismuth phosphate precipitate, thereby increasing the amount of fission products removed from solution.

  16. Thermal stability of fission gas bubble superlattice in irradiated U–10Mo fuel

    SciTech Connect

    Gan, J.; Keiser, D. D.; Miller, B. D.; Robinson, A. B.; Wachs, D. M.; Meyer, M. K.

    2015-09-01

    To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated U-7Mo dispersion and U-10Mo monolithic fuel plates, a FIB-TEM sample of the irradiated U-10Mo fuel with a local fission density of 3.5×1021 fissions/cm3 was used for an in-situ heating TEM experiment. The temperature of the heating holder was raised at a ramp rate of approximately 10 ºC/min up to ~700 ºC, kept at that temperature for about 34 min, continued to 850 ºC with a reduced rate of 5 ºC/min. The result shows a high thermal stability of the fission gas bubble superlattice. The implication of this observation on the fuel microstructural evolution and performance under irradiation is discussed.

  17. A proposed standard on medical isotope production in fission reactors

    SciTech Connect

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-07-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  18. 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.

  19. Immobilization of fission products in phosphate ceramic waste forms

    SciTech Connect

    Singh, D.; Wagh, A.

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  20. Assessment of selected fission products in the Savannah River Site environment

    SciTech Connect

    Carlton, W.H.; Denham, M.

    1997-04-01

    Most of the radioactivity produced by the operation of a nuclear reactor results from the fission process, during which the nucleus of a fissionable atom (such as 235U) splits into two or more nuclei, which typically are radioactive. The Radionuclide Assessment Program (RAP) has reported on fission products cesium, strontium, iodine, and technetium. Many other radionuclides are produced by the fission process. Releases of several additional fission products that result in dose to the offsite population are discussed in this publication. They are 95Zr, 95Nb, 103Ru, 106Ru, 141Ce, and 144Ce. This document will discuss the production, release, migration, and dose to humans for each of these selected fission products.

  1. 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

  2. Gas production apparatus

    DOEpatents

    Winsche, Warren E.; Miles, Francis T.; Powell, James R.

    1976-01-01

    This invention relates generally to the production of gases, and more particularly to the production of tritium gas in a reliable long operating lifetime systems that employs solid lithium to overcome the heretofore known problems of material compatibility and corrosion, etc., with liquid metals. The solid lithium is irradiated by neutrons inside low activity means containing a positive (+) pressure gas stream for removing and separating the tritium from the solid lithium, and these means are contained in a low activity shell containing a thermal insulator and a neutron moderator.

  3. Preliminary investigation of a technique to separate fission noble metals from fission product mixtures

    SciTech Connect

    Mellinger, G.B.; Jensen, G.A.

    1982-08-01

    A variation of the gold-ore fire assay technique was examined as a method for recovering Pd, Rh and Ru from fission products. The mixture of fission product oxides is combined with glass-forming chemicals, a metal oxide such as PbO (scavenging agent), and a reducing agent such as charcoal. When this mixture is melted, a metal button is formed which extracts the noble metals. The remainder cools to form a glass for nuclear waste storage. Recovery depended only on reduction of the scavenger oxide to metal. When such reduction was achieved, no difference in noble metal recovery efficiency was found among the scavengers studied (PbO, SnO, CuO, Bi/sub 2/O/sub 3/, Sb/sub 2/O/sub 3/). Not all reducing agents studied, however, were able to reduce all scavenger oxides to metal. Only graphite would reduce SnO and CuO and allow noble metal recovery. The scavenger oxides Sb/sub 2/O/sub 3/, Bi/sub 2/O/sub 3/, and PbO, however, were reduced by all of the reducing agents tested. Similar noble metal recovery was found with each. Lead oxide was found to be the most promising of the potential scavengers. It was reduced by all of the reducing agents tested, and its higher density may facilitate the separation. Use of lead oxide also appeared to have no deterimental effect on the glass quality. Charcoal was identified as the preferred reducing agent. As long as a separable metal phase was formed in the melt, noble metal recovery was not dependent on the amount of reducing agent and scavenger oxide. High glass viscosities inhibited separation of the molten scavenger, while low viscosities allowed volatile loss of RuO/sub 4/. A viscosity of approx. 20 poise at the processing temperature offered a good compromise between scavenger separation and Ru recovery. Glasses in which PbO was used as the scavenging agent were homogeneous in appearance. Resistance to leaching was close to that of certain waste glasses reported in the literature. 12 figures. 7 tables.

  4. Most probable charge of fission products in 24 MeV proton induced fission of {sup 238}U

    SciTech Connect

    Kudo, H.; Maruyama, M.; Tanikawa, M.; Shinozuka, T.; Fujioka, M.

    1998-01-01

    The charge distributions of fission products in 24 MeV proton-induced fission of {sup 238}U were measured by the use of an ion-guide isotope separator on line. The most probable charge (Z{sub p}) of the charge distribution was discussed in view of the charge polarization in the fission process. It was found that Z{sub p} mainly lies on the proton-rich side in the light mass region and on the proton-deficient side in the heavy mass region compared with the postulate of the unchanged charge distribution. The charge polarization was examined with respect to production Q values. {copyright} {ital 1998} {ital The American Physical Society}

  5. Experimental Measurements of Short-Lived Fission Products from Uranium, Neptunium, Plutonium and Americium

    SciTech Connect

    Metz, Lori A.; Payne, Rosara F.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Pierson, Bruce D.

    2009-11-01

    Fission yields are especially well characterized for long-lived fission products. Modeling techniques incorporate numerous assumptions and can be used to deduce information about the distribution of short-lived fission products. This work is an attempt to gather experimental (model-independent) data on the short-lived fission products. Fissile isotopes of uranium, neptunium, plutonium and americium were irradiated under pulse conditions at the Washington State University 1 MW TRIGA reactor to achieve ~108 fissions. The samples were placed on a HPGe (high purity germanium) detector to begin counting in less than 3 minutes post irradiation. The samples were counted for various time intervals ranging from 5 minutes to 1 hour. The data was then analyzed to determine which radionuclides could be quantified and compared to the published fission yield data.

  6. 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.

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

    DOEpatents

    Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

    1989-01-01

    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.

  8. Fission Product Yields from Fission Spectrum n+{sup 239}Pu for ENDF/B-VII.1

    SciTech Connect

    Chadwick, M.B.; Kawano, T.; Barr, D.W.; Mac Innes, M.R.; Kahler, A.C.; Graves, T.; Selby, H.; Burns, C.J.; Inkret, W.C.; Keksis, A.L.; Lestone, J.P.; Sierk, A.J.; Talou, P.

    2010-12-15

    We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small - especially for {sup 99}Mo - we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for {sup 95}Zr, {sup 140}Ba, {sup 144}Ce), but are larger for {sup 99}Mo (4%-relative) and {sup 147}Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the {sup 147}Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends

  9. Fission product scrubbing system for a nuclear reactor

    SciTech Connect

    Leach, D.S.

    1986-09-09

    A fission product scrubbing system is described for a nuclear reactor including a containment building defining a containment space for accommodating reactor components, comprising (a) means defining a water tank in the containment building; (b) a dividing wall extending into the water tank for separating the water tank into a first and a second compartment; (c) means defining a collection plenum normally hermetically sealed from the containment space and the environment externally of the containment building; (d) means defining a communication passage in the dividing wall underneath the water level in the first and second compartments for maintaining communication between the water stored in the first and second compartments; (e) a standpipe extending from the containment space into the second compartment; (f) a vent pipe extending from the collection plenum into the environment externally of the containment building; and (g) a rupture disc mounted in the vent pipe for normally blocking communication between the collection plenum and the environment.

  10. ZIRCONIUM AND FISSION PRODUCT MANAGEMENT IN THE ALSEP PROCESS

    SciTech Connect

    Lumetta, Gregg J.; Carter, Jennifer C.; Niver, Cynthia M.; Gelis, Artem V.

    2013-09-29

    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. 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 HNO3, so these components can easily be routed to the high-level waste raffinate.

  11. SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

    DOEpatents

    Maddock, A.G.; Booth, A.H.

    1960-09-13

    Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

  12. Behavior of fission product tellurium under severe accident conditions

    SciTech Connect

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

    1986-01-01

    Fission product release tests at Oak Ridge National Laboratory (ORNL) have provided new experimental data that help characterize the behavior of tellurium under severe light-water reactor (LWR) accident conditions. The release of tellurium from the fuel rods is dependent upon the rate and extent of cladding oxidation. Tellurium has been found to be considerably retained by metallic Zircaloy cladding at test temperatures up to 1975/sup 0/C. The results indicate that the tellurium is bound by the Zircaloy cladding as zirconium telluride, but once the available zirconium metal is oxidized by the steam, tellurium is released in favor of continued zirconium oxide formation. The collection behavior of the released tellurium indicates that it is probably released from the fuel rods as SnTe and CsTe, rather than as elemental tellurium.

  13. ORNL studies of fission product release under LWR accident conditions

    SciTech Connect

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

    1991-01-01

    High burnup Zircaloy-clad UO{sub 2} fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction.

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

    SciTech Connect

    Mertz, C.; Chamberlain, D.B.; Chen, L.; Conner, C.; Vandegrift, G.F.; Drockelman, D.; Kaminski, M.; Landsberger, S.; Stubbins, J.

    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.

  15. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

    DOEpatents

    Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

    1962-11-13

    A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

  16. Fission product source term research at Oak Ridge National Laboratory. [PWR; BWR

    SciTech Connect

    Malinauskas, A.P.

    1985-01-01

    The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, accident sequence progression, etc., will be discussed.

  17. Fission product release and fuel cladding interaction in severe-accident tests of LWR fuel

    SciTech Connect

    Strain, R.V.; Osborne, M.F.

    1983-11-01

    The examination of these samples indicated a correlation between the posttest fuel microstructure and the fission product release during the test. As expected, structural changes in the fuel and fission product release increased with test temperature. The effect of steam flow rate, which controls the extent of cladding oxidation, however, was less clear. The amount of fuel-cladding reaction and liquefaction was greatest in the test with a low steam flow rate, which was also the highest temperature test. Other data indicate, however, that extensive fuel-cladding reaction and liquefaction would be expected at approx. 1700/sup 0/C with reduced steam flow rate (i.e., with reduced oxidation). The similar gas release values and fuel microstructures for the 1700 and 2000/sup 0/C test are somewhat surprising, but may indicate the influence of the steam conditions on gas release as well as on fuel-cladding reaction. The extent of fuel-cladding interaction in these tests, and the resulting intermediate phases, appear to be consistent with the observations of Hofmann and Kerwin-Peck.

  18. Neutron Cross Section Covariances for Structural Materials and Fission Products

    NASA Astrophysics Data System (ADS)

    Hoblit, S.; Cho, Y.-S.; Herman, M.; Mattoon, C. M.; Mughabghab, S. F.; Obložinský, P.; Pigni, M. T.; Sonzogni, A. A.

    2011-12-01

    We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10 eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also 23Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

  19. Neutron Cross Section Covariances for Structural Materials and Fission Products

    SciTech Connect

    Hoblit, S.; Hoblit,S.; Cho,Y.-S.; Herman,M.; Mattoon,C.M.; Mughabghab,S.F.; Oblozinsky,P.; Pigni,M.T.; Sonzogni,A.A.

    2011-12-01

    We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10{sup -5} eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also {sup 23}Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

  20. Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

    SciTech Connect

    Andersson, Anders David Ragnar; Pastore, Giovanni; Liu, Xiang-Yang; Perriot, Romain Thibault; Tonks, Michael; Stanek, Christopher Richard

    2014-11-07

    This report summarizes the development of new fission gas diffusion models from lower length scale simulations and assessment of these models in terms of annealing experiments and fission gas release simulations using the BISON fuel performance code. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations, continuum models for diffusion of xenon (Xe) in UO2 were derived for both intrinsic conditions and under irradiation. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.

  1. Assessment of fission product yields data needs in nuclear reactor applications

    SciTech Connect

    Kern, K.; Becker, M.; Broeders, C.

    2012-07-01

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

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

    SciTech Connect

    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.

  3. Measurement of fission products yields in the quasi-mono-energetic neutron-induced fission of 232Th

    NASA Astrophysics Data System (ADS)

    Naik, H.; Mukherji, Sadhana; Suryanarayana, S. V.; Jagadeesan, K. C.; Thakare, S. V.; Sharma, S. C.

    2016-08-01

    The cumulative yields of various fission products in the 232Th(n, f) reaction at average neutron energies of 5.42, 7.75, 9.35 and 12.53 MeV have been determined by using an off-line γ-ray spectrometric technique. The neutron beam was produced from the 7Li(p, n) reaction by using the proton energies of 7.8, 12, 16 and 20 MeV. The mass chain yields were obtained from the cumulative fission yields by using the charge distribution correction of medium energy fission. The fine structure in the mass yield distribution was interpreted from the point of nuclear structure effect. On the other hand, the higher yield around mass number 133-134 and 143-144 as well as their complementary products were explained based on the standard I and standard II asymmetric mode of fission. From the mass yield data, the average value of light mass (), heavy mass (), the average number of neutrons (< ν >) and the peak-to-valley (P / V) ratios at different neutron energies of present work and literature data were obtained in the 232Th(n, f) reaction. The different parameters of the mass yield distribution in the 232Th(n, f) reaction were compared with the similar data in the 232Th(γ, f) reaction at comparable excitation energy and a surprising difference was observed.

  4. HTR 2014 Paper - Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise P. Collin

    2001-10-01

    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 1800°C 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 an overall over-prediction of the fractional release of these fission products, which is largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. Correction factors to these diffusivities were assessed for silver and cesium in order to enable a better match between the modeling predictions and the safety testing results. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. In the case of krypton, all the coating layers are partly retentive and the available data did not allow to determine their respective retention powers, hence preventing to derive any correction factors.

  5. 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

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

    SciTech Connect

    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.

    2010-12-15

    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 {sup 99}Mo, {sup 95}Zr, {sup 137}Cs, {sup 140}Ba, {sup 141,143}Ce, and {sup 147}Nd. Modest incident-energy dependence exists for the {sup 147}Nd 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 {approx}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

  7. High-Resolution Correlated Fission Product Measurements of 235U (nth , f) with SPIDER

    NASA Astrophysics Data System (ADS)

    Shields, Dan; Spider Team

    2015-10-01

    The SPIDER detector (SPectrometer for Ion DEtermination in fission Research) has obtained high-resolution, moderate-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 (ν), and total kinetic energy (TKE). These data will be some of the first of their kind available to nuclear data evaluations. An overview of the SPIDER detector, analytical method, and preliminary results for 235U (nth , f) will be presented. LA-UR-15-20130 This work benefited from the use of the LANSCE accelerator facility and was performed under the auspices of the US Department of Energy by Los Alamos Security, LLC under Contract DE-AC52-06NA25396.

  8. Short-lived fission product measurements from >0.1 MeV neutron-induced fission using boron carbide.

    SciTech Connect

    Finn, Erin C.; Metz, Lori A.; Greenwood, Lawrence R.; Pierson, Bruce D.; Friese, Judah I.; Kephart, Rosara F.; Kephart, Jeremy D.

    2012-02-01

    A boron carbide shield was designed, custom fabricated, and used to create a fast fission energy neutron spectrum. The fissionable isotopes 233, 235, 238U, 237Np, and 239Pu were separately placed inside of this shield and irradiated under pulsed conditions at the Washington State University 1 MW TRIGA reactor. A unique set of fission product gamma spectra were collected at short times (4 minutes to 1 week) post-fission. Gamma spectra were collected on single-crystal high purity germanium detectors and on Pacific Northwest National Laboratory's (PNNL's) Direct Simultaneous Measurement (DSM) system composed of HPGe detectors connected in coincidence. This work defines the experimental methods used to produce and collect the gamma data, and demonstrates the validity of the measurements. It is important to fully document this information so the data can be used with high confidence for the advancement of nuclear science and non-proliferation applications. The gamma spectra collected in these and other experiments will be made publicly available at https://spcollab.pnl.gov/sites/gammadata or via the link at http://rdnsgroup.pnl.gov. A revised version of this publication will be posted with the data to make the experimental details available to those using the data.

  9. Fission gas release from oxide fuels at high burnups (AWBA development program)

    SciTech Connect

    Dollins, C.C.

    1981-02-01

    The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement.

  10. Critical temperature for the nuclear liquid-gas phase transition (from multifragmentation and fission)

    SciTech Connect

    Karnaukhov, V. A.; Oeschler, H.; Budzanowski, A.; Avdeyev, S. P.; Botvina, A. S.; Cherepanov, E. A.; Karcz, W.; Kirakosyan, V. V.; Rukoyatkin, P. A.; Skwirczynska, I.; Norbeck, E.

    2008-12-15

    Critical temperature T{sub c} for the nuclear liquid-gas phase transition is estimated from both the multifragmentation and fission data. In the first case, the critical temperature is obtained by analysis of the intermediate-mass-fragment yields in p(8.1 GeV) + Au collisions within the statistical model of multifragmentation. In the second case, the experimental fission probability for excited {sup 188}Os is compared with the calculated one with T{sub c} as a free parameter. It is concluded for both cases that the critical temperature is higher than 15 MeV.

  11. Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

    SciTech Connect

    Pierson, Bruce D.; Finn, Erin C.; Friese, Judah I.; Greenwood, Lawrence R.; Kephart, Jeremy D.; Kephart, Rosara F.; Metz, Lori A.

    2013-03-01

    Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University (WSU) Nuclear Radiation Center’s 1MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective, the ORTEC Detective, and the Canberra Falcon. Canberra’s GENIE-2000 software was used to analyze the spectral data collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.

  12. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Terrestrial and Water Ecosystems

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    A large number of studies and models were established to explain the fission products (FP) behavior within terrestrial and water ecosystems, but a number of behaviors were non understandable, which always attributed to unknown reasons. According to DAB hypothesis, almost all fission products behaviors in terrestrial and water ecosystems could be interpreted in a wide coincidence. The gab between former models predictions, and field behavior of fission products after accidents like Chernobyl have been explained. DAB represents a tool to reduce radio-phobia as well as radiation protection expenses. (author)

  13. SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS

    DOEpatents

    Schubert, J.

    1960-05-24

    A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

  14. Zirconium and fission product management in the ALSEP process

    SciTech Connect

    Lumetta, G.J.; Carter, J.C.; Niver, C.M.

    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)

  15. 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.

    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.

  16. 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.

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

    PubMed

    Bihari, Árpád; Dezső, Zoltán; Bujtás, Tibor; Manga, László; Lencsés, András; Dombóvári, Péter; Csige, István; Ranga, Tibor; Mogyorósi, Magdolna; Veres, Mihály

    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 Bátaapáti (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 2.1±0.1 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

  18. 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

  19. Determination of {sup 140}La fission product interference factor for INAA

    SciTech Connect

    Ribeiro Jr, Iberê S.; Genezini, Frederico A.; Saiki, Mitiko; Zahn, Guilherme S.

    2014-11-11

    Instrumental Neutron Activation Analysis (INAA) is a technique widely used to determine the concentration of several elements in several kinds of matrices. However if the sample of interest has higher relative uranium concentration the obtained results can be interfered by the uranium fission products. One of these cases that is affected by interference due to U fission is the {sup 140}La, because this radioisotope used in INAA for the determination of concentration the La is also produced by the {sup −}β of {sup 140}Ba, an uranium fission product. The {sup 140}La interference factor was studied in this work and a factor to describe its time dependence was obtained.

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

    SciTech Connect

    Honaiser, Eduardo; Anghaie, Samim

    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)

  1. The LANL C-NR counting room and fission product yields

    SciTech Connect

    Jackman, Kevin Richard

    2015-09-21

    This PowerPoint presentation focused on the following areas: LANL C-NR counting room; Fission product yields; Los Alamos Neutron wheel experiments; Recent experiments ad NCERC; and Post-detonation nuclear forensics

  2. ARSENATE CARRIER PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM NEUTRON IRRADIATED URANIUM AND RADIOACTIVE FISSION PRODUCTS

    DOEpatents

    Thompson, S.G.; Miller, D.R.; James, R.A.

    1961-06-20

    A process is described for precipitating Pu from an aqueous solution as the arsenate, either per se or on a bismuth arsenate carrier, whereby a separation from uranium and fission products, if present in solution, is accomplished.

  3. Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

    SciTech Connect

    White, J.R.; Burns, T.J.

    1980-02-01

    The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.

  4. Investigation of Inconsistent ENDF/B-VII.1 Independent and Cumulative Fission Product Yields with Proposed Revisions

    NASA Astrophysics Data System (ADS)

    Pigni, M. T.; Francis, M. W.; Gauld, I. C.

    2015-01-01

    A recent implementation of ENDF/B-VII.1 independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear schemes in the decay sub-library that are 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 do not agree with the cumulative fission yields in the library as well as with experimental measurements. To address these issues, 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 compare the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. Another 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 for stable and long-lived fission products. 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.

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

    SciTech Connect

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

    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.

  6. Investigation of Inconsistent ENDF/B-VII.1 Independent and Cumulative Fission Product Yields with Proposed Revisions

    SciTech Connect

    Pigni, M.T. Francis, M.W.; Gauld, I.C.

    2015-01-15

    A recent implementation of ENDF/B-VII.1 independent fission product yields and nuclear decay data identified inconsistencies in the data caused by the use of updated nuclear schemes in the decay sub-library that are 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 do not agree with the cumulative fission yields in the library as well as with experimental measurements. To address these issues, a comprehensive set of independent fission product yields was generated for thermal and fission spectrum neutron-induced fission for {sup 235,238}U and {sup 239,241}Pu 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 compare the calculated fission product inventories with experimentally measured inventories, with particular attention given to the noble gases. Another 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 for stable and long-lived fission products. 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.

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

    NASA Astrophysics Data System (ADS)

    Bunakov, V. E.; Kadmensky, S. G.; Kadmensky, S. S.

    2008-11-01

    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 nonevaporative 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.

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

    SciTech Connect

    Bunakov, V. E. Kadmensky, S. G. Kadmensky, S. S.

    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.

  9. RARE-EARTH METAL FISSION PRODUCTS FROM LIQUID U-Bi

    DOEpatents

    Wiswall, R.H.

    1960-05-10

    Fission product metals can be removed from solution in liquid bismuth without removal of an appreciable quantity of uranium by contacting the liquid metal solution with fused halides, as for example, the halides of sodium, potassium, and lithium and by adding to the contacted phases a quantity of a halide which is unstable relative to the halides of the fission products, a specific unstable halide being MgCl/sub 3/.

  10. Performance of RELAP/SCDAPSIM Code on Fission Products Transport Prediction

    SciTech Connect

    Honaiser, Eduardo

    2006-07-01

    Fission product transport in the piping system of primary circuits is an important area of study in field of the severe accidents. Fission product transport comprises all phenomenon occurring from the nuclear core to the containment release site. Once released in the flow channels, fission products can condense on the piping walls, nucleate aerosols, which can agglomerate and/or deposit on the piping walls. The phenomenology occurs in a steam-hydrogen convective environment. A model (FPTRAN) was developed for the program RELAP/SCDAPSIM that calculates all phenomenon related to the fission product transport through the piping system. The model solves a set of differential equations. The coefficients in these equations represent the processes at which several states change among them. The processes considered were vapor adsorption and condensation on the piping walls, aerosol formation and growth (condensation and agglomeration), and aerosol deposition. The model also controls the aerosol particle size distribution. The PHEBUS experiments compose the most complete experimental program ever conducted for the understanding of fission product behavior in Reactor Cooling System and containment. It employs a reactor to generate fission products, which are transported through a scaled piping system simulating the primary circuit of a pressurized water reactor (PWR). Along the piping system, several instruments are installed to measure the amount of fission products deposited and their states. This paper describes the modeling of the experiment Phebus FPT-01 using RELAP/SCDAPSIM and compares simulation and experimental results to assess the performance of the FPTRAN module on the fission products transport prediction. These results can be considered satisfactory, except for iodine. This inconsistency of iodine is probably due to an incorrect chemical form assumed for iodine. (author)

  11. Mass spectrometric study of the release of volatile fission products from irradiated LWR fuel

    SciTech Connect

    Johnson, I.; Steidl, D.V.; Johnson, C.E.

    1984-01-01

    The objective of these studies is to experimentally determine the chemical form and the rate of release of volatile fission product species from defected irradiated LWR reactor fuel pins. After release from the defected fuel pin the gaseous species immediately enters the ionizer of a quadrupole mass spectrometer thus ensuring that their chemical form is not likely to be changed prior to identification and measurement. These studies differ from prior studies in that: (1) the chemical form of the volatile fission products will be determined; and (2) the detection and measurement method does not depend on the radioactivity of the fission product element. Information on the chemical form of the released fission product species will enable a more accurate description of their transport and reaction in the primary system. These studies are also expected to yield information on the reaction of fission products after release from the fuel oxide with the zircaloy cladding. The results of these studies are expected to increase the understanding of the first step in the release of fission products by irradiated fuel and therefore help in the accurate prediction of source terms.

  12. 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.

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

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    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.

  14. Fission signal detection using helium-4 gas fast neutron scintillation detectors

    SciTech Connect

    Lewis, J. M. Kelley, R. P.; Jordan, K. A.; Murer, D.

    2014-07-07

    We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure {sup 4}He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the {sup 4}He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

  15. Simulating γ-γ coincidences of β-delayed γ-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 γ-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 γ-ray intensities from nuclear databases are inputs into a GEANT4 simulation to create expected γ-ray spectra from irradiated 235U. The simulations include not only isotropically emitted γ-rays but also γ-γ cascades from certain fission products, emitted with their appropriate angular correlations. Here γ singles spectra as well as γ-γ 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 γ-γ cascades in fission product decays to determine isotopic content. Analyzing experimental and simulated γ-γ 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 γ singles spectra.

  16. Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

    SciTech Connect

    Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

    2015-01-01

    The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δkeff (ISG-8, Rev. 3, Recommendation 4).

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

    DOE PAGESBeta

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2016-04-07

    Safety tests were conducted on 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 1800 °C 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 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compactsmore » containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed

  18. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    NASA Astrophysics Data System (ADS)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  19. 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.

  20. 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.

  1. Fission yield measurements at IGISOL

    NASA Astrophysics Data System (ADS)

    Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

    2016-06-01

    The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  2. Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges

    SciTech Connect

    Bibler, N

    2005-04-05

    The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations of the noble metals in HLW sludges are measured by using mass spectroscopy to determine concentrations of the isotopes that comprise each noble metal. For example, the noble metal Ru is comprised

  3. Fission Product Removal From Spent Oxide Fuel By Head-End Processing

    SciTech Connect

    B. R. Westphal; K. J. Bateman; R. P. Lind; K. L. Howden; G. D. Del Cul

    2005-10-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. Experiments have been performed with irradiated oxide fuel to evaluate the removal of fission products. During these experiments, operating parameters such as temperature and pressure have been varied to discern their effects on the behavior of specific fission products. In general, the extent of removal increases with increasing operating temperature and decreasing pressure. Removal efficiencies as high as 98% have been achieved during testing. Given the results of testing, an explanation of the likely fission product species being removed during the test program is also provided. In addition, experiments have been performed with other oxidative gases (steam and ozone) on surrogates to determine their potential benefit for removal of fission products.

  4. 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.

  5. Fission product retention in newly discovered organic-rich natural fission reactors at Oklo and Bangombe, Gabon

    SciTech Connect

    Nagy, B.; Rigali, M.J. )

    1993-01-01

    The discovery of naturally occurring fission reactors in the rock strata of the Paleoproterozoic Francevillian Basin in the Republic of Gabon in equatorial West Africa led to several programs to define migration and/or retention of uranium and fissiogenic isotopes from/in the natural reactor zones. Although much understanding has been gained, new insight is needed regarding the chemical and physical parameters that control movement and retention of fission products over almost two billion years from/in the natural reactors. Seventeen known natural fission reactors sustained criticality for 0.1 to 1 million years in hydrothermally altered sedimentary rocks 1968 +/- 50 million years ago. These natural nuclear reactors attained criticality because of high concentrations of uranium in small pockets in uranium ores, the lack of neutron poisons, and because at the time they reached criticality, the abundance of [sup 235]U was five times greater than it is today. Water acted as a moderator, and temperature in the natural reactors was between 160 and 360[degrees]C. Both the uranium-rich pockets and the uranium ore bodies in which these pockets are located were formed when aqueous solutions moving through highly fractured zones in the Francevillian sedimentary rocks met organic-rich sediments. This resulted in the reduction of U(VI) in the dissolved uranyl ions to U(IV), causing the precipitation of pitchblende and uraninite. It has been proposed that between 2.2 and 1.9 billion years ago, the earth's atmosphere experienced a remarkable temporary rise in O[sub 2] content; this event may account for the uranium-bearing, oxidizing aqueous solutions in the Francevillian rocks.

  6. 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.

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

    SciTech Connect

    Thompton, J.L.; Gilmore, J.S. )

    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.

  8. Influence of fission products on ruthenium oxidation and transport in air ingress nuclear accidents

    NASA Astrophysics Data System (ADS)

    Vér, N.; Matus, L.; Kunstár, M.; Osán, J.; Hózer, Z.; Pintér, A.

    2010-01-01

    In separate effect tests at 1000-1200 °C Ru oxidation rate and content of Ru in escaping air flow have been studied with special emphasis on effects of other fission product elements on the Ru oxidation and transport. The results showed that in the decreasing temperature section (1100-600 °C) most of the RuO3 and RuO4 (≈95%) decomposed and formed RuO2 crystals; while the partial pressure of RuO4 in the escaping air was in the range of 10-6 bar. The re-evaporation of deposited RuO2 resulted in about 10-6 bar partial pressure in the outlet gas as well. Measurements demonstrated the importance of surface quality in the decreasing temperature area on the heterogeneous phase decomposition of ruthenium oxides to RuO2. On the other hand water or molybdenum oxide vapour in air appears to decrease the surface catalyzed decomposition of RuOx to RuO2 and increases RuO4 concentration in the escaping air. High temperature reaction with caesium changed the form of the released ruthenium and caused a time delay in appearance of maximum concentration of ruthenium oxides in the ambient temperature escaping gas, while reaction with barium and rare earth oxides extended Ru escape from the high temperature area.

  9. Precise ruthenium fission product isotopic analysis using dynamic reaction cell inductively coupled plasma mass spectrometry (DRC-ICP-MS)

    SciTech Connect

    Brown, Christopher F.; Dresel, P. Evan; Geiszler, Keith N.; Farmer, Orville T.

    2006-05-09

    99Tc is a subsurface contaminant of interest at numerous federal, industrial, and international facilities. However, as a mono-isotopic fission product, 99Tc lacks the ability to be used as a signature to differentiate between the different waste disposal pathways that could have contributed to subsurface contamination at these facilities. Ruthenium fission-product isotopes are attractive analogues for the characterization of 99Tc sources because of their direct similarity to technetium with regard to subsurface mobility, and their large fission yields and low natural background concentrations. We developed an inductively coupled plasma mass spectrometry (ICP-MS) method capable of measuring ruthenium isotopes in groundwater samples and extracts of vadose zone sediments. Samples were analyzed directly on a Perkin Elmer ELAN DRC II ICP-MS after a single pass through a 1-ml bed volume of Dowex AG 50W-X8 100-200 mesh cation exchange resin. Precise ruthenium isotopic ratio measurements were achieved using a low-flow Meinhard-type nebulizer and long sample acquisition times (150,000 ms). Relative standard deviations of triplicate replicates were maintained at less than 0.5% when the total ruthenium solution concentration was 0.1 ng/ml or higher. Further work was performed to minimize the impact caused by mass interferences using the dynamic reaction cell (DRC) with O2 as the reaction gas. The aqueous concentrations of 96Mo and 96Zr were reduced by more than 99.7% in the reaction cell prior to injection of the sample into the mass analyzer quadrupole. The DRC was used in combination with stable-mass correction to quantitatively analyze samples containing up to 2-orders of magnitude more zirconium and molybdenum than ruthenium. The analytical approach documented herein provides an efficient and cost-effective way to precisely measure ruthenium isotopes and quantitate total ruthenium (natural vs. fission-product) in aqueous matrixes.

  10. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  11. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 3: Fission-Product Transport and Dose PIRTs

    SciTech Connect

    Morris, Robert Noel

    2008-03-01

    This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into the

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

    SciTech Connect

    Katakura, J. ); England, T.R. )

    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.

  13. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    NASA Astrophysics Data System (ADS)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  14. PROCESS USING BISMUTH PHOSPHATE AS A CARRIER PRECIPITATE FOR FISSION PRODUCTS AND PLUTONIUM VALUES

    DOEpatents

    Finzel, T.G.

    1959-03-10

    A process is described for separating plutonium from fission products carried therewith when plutonium in the reduced oxidation state is removed from a nitric acid solution of irradiated uranium by means of bismuth phosphate as a carrier precipitate. The bismuth phosphate carrier precipitate is dissolved by treatment with nitric acid and the plutonium therein is oxidized to the hexavalent oxidation state by means of potassium dichromate. Separation of the plutonium from the fission products is accomplished by again precipitating bismuth phosphate and removing the precipitate which now carries the fission products and a small percentage of the plutonium present. The amount of plutonium carried in this last step may be minimized by addition of sodium fluoride, so as to make the solution 0.03N in NaF, prior to the oxidation and prccipitation step.

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

    SciTech Connect

    Xiao, Haiyan; Zhang, Yanwen; Weber, William J

    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.

  16. Results of fission products β decay properties measurement performed with a total absorption spectrometer

    NASA Astrophysics Data System (ADS)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2014-03-01

    β-decay properties of fission products are very important for applied reactor physics, for instance to estimate the decay heat released immediately after the reactor shutdown and to estimate the bar ν flux emitted. An accurate estimation of the decay heat and the bar ν emitted flux from reactors, are necessary for purposes such as reactors operation safety and non-proliferation. In order to improve the precision in the prediction for these quantities, the bias due to the Pandemonium effect affecting some important fission product data has to be corrected. New measurements of fission products β-decay, not sensitive to this effect, have been performed with a Total Absorption Spectrometer (TAS) at the JYFL facility of Jyväskylä. An overview of the TAS technique and first results from the 2009 campaign will be presented.

  17. Modeling the influence of bubble pressure on grain boundary separation and fission gas release

    SciTech Connect

    Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

    2014-09-01

    Grain boundary (GB) separation as a mechanism for fission gas release (FGR), complementary to gas bubble interlinkage, has been experimentally observed in irradiated light water reactor fuel. However there has been limited effort to develop physics-based models incorporating this mechanism for the analysis of FGR. In this work, a computational study is carried out to investigate GB separation in UO2 fuel under the effect of gas bubble pressure and hydrostatic stress. A non-dimensional stress intensity factor formula is obtained through 2D axisymmetric analyses considering lenticular bubbles and Mode-I crack growth. The obtained functional form can be used in higher length-scale models to estimate the contribution of GB separation to FGR.

  18. A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Beneš, O.; Konings, R. J. M.

    2015-10-01

    The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas-gas and gas-aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

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

    NASA Astrophysics Data System (ADS)

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

    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 UO 2 fuel. The results show that in the burn-up range investigated, SBR MOX fuel and MIMAS MOX fuel perform similarly, releasing up to about 1% of the fission gas inventory. Comparisons with the Halden Criterion show that SBR MOX has the same release threshold as UO 2 and this suggests that the mechanisms of release in the two fuels are similar. This is further supported by calculations made with the ENIGMA fuel performance code. It is concluded that the apparent differences in fission gas release between SBR MOX and UO 2 fuel, at least in the early stages of release, can be explained by the higher temperatures experienced by MOX fuel.

  20. 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.

  1. Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

    SciTech Connect

    Wichner, R.P.

    1983-01-01

    The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rates and transport in the reactor vessel during LWR core-melt accidents were studied. The Nuclear Safety Pilot Plant is dedicated to developing an expanded data base on the behavior of aerosols generated during a severe accident.

  2. 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.

  3. Integrated separation scheme for measuring a suite of fission and activation products from a fresh mixed fission and activation product sample

    SciTech Connect

    Morley, Shannon M.; Seiner, Brienne N.; Finn, Erin C.; Greenwood, Lawrence R.; Smith, Steven C.; Gregory, Stephanie J.; Haney, Morgan M.; Lucas, Dawn D.; Arrigo, Leah M.; Beacham, Tere A.; Swearingen, Kevin J.; Friese, Judah I.; Douglas, Matthew; Metz, Lori A.

    2015-05-01

    Mixed fission and activation materials resulting from various nuclear processes and events contain a wide range of isotopes for analysis spanning almost the entire periodic table. In some applications such as environmental monitoring, nuclear waste management, and national security a very limited amount of material is available for analysis and characterization so an integrated analysis scheme is needed to measure multiple radionuclides from one sample. This work describes the production of a complex synthetic sample containing fission products, activation products, and irradiated soil and determines the percent recovery of select isotopes through the integrated chemical separation scheme. Results were determined using gamma energy analysis of separated fractions and demonstrate high yields of Ag (76 ± 6%), Au (94 ± 7%), Cd (59 ± 2%), Co (93 ± 5%), Cs (88 ± 3%), Fe (62 ± 1%), Mn (70 ± 7%), Np (65 ± 5%), Sr (73 ± 2%) and Zn (72 ± 3%). Lower yields (< 25%) were measured for Ga, Ir, Sc, and W. Based on the results of this experiment, a complex synthetic sample can be prepared with low atom/fission ratios and isotopes of interest accurately and precisely measured following an integrated chemical separation method.

  4. HTR-2014 Paper Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2001-10-01

    The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of fission products silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of fission products from fuel compacts and fuel particles, and retention of fission products in the compacts outside of the SiC layer. PARFUME-predicted fractional release of these fission products was determined and compared to the PIE measurements. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed silicon carbide (SiC) layers, the over-prediction is by a factor of about two, corresponding to an over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of about 100. For intact particles, whose release is much lower, the over-prediction is by an average of about an order of magnitude, which could additionally be attributed to an over-estimated diffusivity in SiC by about 30%. The release of strontium from intact particles is also over-estimated by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Furthermore, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the

  5. Concentration-triggered fission product release from zirconia: consequences for nuclear safety

    NASA Astrophysics Data System (ADS)

    Gentils, A.; Thomé, L.; Jagielski, J.; Garrido, F.

    2002-02-01

    Crystalline oxide ceramics, more particularly zirconia and spinel, are promising matrices for plutonium and minor actinide transmutation. An important issue concerning these materials is the investigation of their ability to confine radiotoxic elements resulting from the fission of actinides. This letter reports the study of the release, upon annealing or irradiation at high temperature, of one of the most toxic fission product (Cs) in zirconia. The foreign species are introduced by ion implantation and the release is studied by Rutherford backscattering experiments. The results emphasize the decisive influence of the fission product concentration on the release properties. The Cs mobility in zirconia is strongly increased when the impurity concentration exceeds a threshold of the order of a few atomic per cent. Irradiation with medium-energy heavy ions is shown to enhance Cs outdiffusion with respect to annealing at the same temperature.

  6. Experiments on the high-temperature behaviour of neutron-irradiated uranium dioxide and fission products, volume 8, number 1

    NASA Astrophysics Data System (ADS)

    Tanke, R. H. J.

    The release rate of fission products from overheated UO2, the chemical form of these fission products, and the transport mechanism inside the nuclear fuel are determined. UO spheres of approximately 1 mm diameter, irradiated in a high-flux reactor were used for the experiments. The chemical forms of the particles released from the spheres during evaporation were determined by mass spectrometry and the release rate of the mission products was determined by gamma spectrometry. A gamma topographer was developed to determine the change with temperature in the three dimensional distribution of radioactive fission products in the spheres. No clear relationship between the stoichiometry of the spheres and uranium consumption were shown. A diffusion model was used to determine the activation energy for the diffusion of fission products. It is concluded that the microstructure of the nuclear fuel greatly affects the number of free oxygen atoms, the release rate and the chemical form of the fission products. The evaporation of the UO2 matrix is the main mechanism for the release of all fission products at temperatures above 2300 K. Barium can be as volatile as iodine. Niobium and lanthenum can be volatile. Molecular combinations of the fission products, iodine, cesium and tellurium, are highly unlikely to be present inside the fuel. Barium and nobium may form compounds with oxygen and are then released as simple oxides. Fission products are released from overheated UO2 or as oxides. A new model is proposed for describing the behavior of oxygen in irradiated nuclear fuel.

  7. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    SciTech Connect

    Asner, David M.; Burns, Kimberly A.; Campbell, Luke W.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wood, Lynn S.; Wootan, David W.

    2015-03-01

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  8. 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.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    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 β and γ 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

  9. 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…

  10. 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.

  11. 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...

  12. SEPARATION OF FISSION PRODUCT VALUES FROM THE HEXAVALENT PLUTONIUM BY CARRIER PRECIPITATION

    DOEpatents

    Davies, T.H.

    1959-12-15

    An improved precipitation of fission products on bismuth phosphate from an aqueous mineral acid solution also containing hexavalent plutonium by incorporating, prior to bismuth phosphate precipitation, from 0.05 to 2.5 grams/ liter of zirconium phosphate, niobium oxide. and/or lanthanum fluoride is described. The plutonium remains in solution.

  13. 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.

  14. 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).

  15. Report on the Behavior of Fission Products in the Co-decontamination Process

    SciTech Connect

    Martin, Leigh Robert; Riddle, Catherine Lynn

    2015-09-30

    This document was prepared to meet FCT level 3 milestone M3FT-15IN0302042, “Generate Zr, Ru, Mo and Tc data for the Co-decontamination Process.” This work was carried out under the auspices of the Lab-Scale Testing of Reference Processes FCT work package. This document reports preliminary work in identifying the behavior of important fission products in a Co-decontamination flowsheet. Current results show that Tc, in the presence of Zr alone, does not behave as the Argonne Model for Universal Solvent Extraction (AMUSE) code would predict. The Tc distribution is reproducibly lower than predicted, with Zr distributions remaining close to the AMUSE code prediction. In addition, it appears there may be an intricate relationship between multiple fission product metals, in different combinations, that will have a direct impact on U, Tc and other important fission products such as Zr, Mo, and Rh. More extensive testing is required to adequately predict flowsheet behavior for these variances within the fission products.

  16. COPAR-FD. Release of Metallic Fission Products from Coated Nuclear Fuel Particles

    SciTech Connect

    Tzung, F.; Richards, M.

    1992-09-01

    COPAR-FD is used to calculate the release of metallic fission products from coated nuclear fuel particles, using a finite-difference solution of the governing partial differential equation. COPAR-FD interfaces with the TRAMP and TRAFIC codes for calculating transport in and release from graphite fuel blocks.

  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. ); Kinard, W.F. . 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. 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.; Kinard, W.F.

    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.

  19. Compilation of Data on Radionuclide Data for Specific Activity, Specific Heat and Fission Product Yields

    SciTech Connect

    Gibbs, A.; Thomason, R.S.

    2000-09-05

    This compilation was undertaken to update the data used in calculation of curie and heat loadings of waste containers in the Solid Waste Management Facility. The data has broad general use and has been cross-checked extensively in order to be of use in the Materials Accountability arena. The fission product cross-sections have been included because they are of use in the Environmental Remediation and Waste Management areas where radionuclides which are not readily detectable need to be calculated from the relative fission yields and material dispersion data.

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

    SciTech Connect

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

    2011-07-01

    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 experimental conditions such as temperature rising up to 150 deg.C and especially, high thermal fluence level up to 3.5 10{sup 19} n.cm{sup 2}, the initial sensor gas speed of sound efficiency measurement was strongly reduced due to the irradiation effects on the piezo-ceramic properties. Nevertheless, by adding a differential signal processing method to the initial data analysis procedure validated before irradiation, the gas resonance peaks were successfully extracted from the output signal. From these data, the molar fractions variations of helium and fission gas were measured from an adapted Virial state equation. Thus, with this sensor, the kinetics of gas release inside fuel rods could be deduced from the in-pile measurements and specific calculations. These data will also give information about nuclear reaction effect on piezo-ceramics sensor under high neutron and gamma flux. (authors)

  1. 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 1800°C 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 1600°C and 1700°C, and at least 10 to 102 at 1800°C. 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 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C 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 1700°C, and 102 to 103 at 1800°C. These

  2. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  3. 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 mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, 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.

  4. 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

    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)

  5. EIA's Natural Gas Production Data

    EIA Publications

    2009-01-01

    This special report examines the stages of natural gas processing from the wellhead to the pipeline network through which the raw product becomes ready for transportation and eventual consumption, and how this sequence is reflected in the data published by the Energy Information Administration (EIA).

  6. 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 an 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.

  7. 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.

  8. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  9. The role of charge and ionic radius on fission product segregation to a model UO2 grain boundary

    NASA Astrophysics Data System (ADS)

    Hong, Minki; Uberuaga, Blas P.; Phillpot, Simon R.; Andersson, David A.; Stanek, Christopher R.; Sinnott, Susan B.

    2013-04-01

    The segregation energies of a range of fission products in UO2 to a Σ5 symmetric tilt grain boundary have been calculated using empirical potentials and their dependency on site, charge, and ionic radius has been determined. Density functional theory calculations provide information about the detailed bonding environment around the segregates. While most of the fission products prefer to reside in sites with large free volume, there are some that form strong bonds with neighboring oxygen ions, and thus prefer sites with high oxygen coordination. This result provides insight into nuclear fuel design to enhance control of fission product retention.

  10. 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.

  11. Enhanced Generic Phase-field Model of Irradiation Materials: Fission Gas Bubble Growth Kinetics in Polycrystalline UO2

    SciTech Connect

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

    2012-05-30

    Experiments show that inter-granular and intra-granular gas bubbles have different growth kinetics which results in heterogeneous gas bubble microstructures in irradiated nuclear fuels. A science-based model predicting the heterogeneous microstructure evolution kinetics is desired, which enables one to study the effect of thermodynamic and kinetic properties of the system on gas bubble microstructure evolution kinetics and morphology, improve the understanding of the formation mechanisms of heterogeneous gas bubble microstructure, and provide the microstructure to macroscale approaches to study their impact on thermo-mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking. In our previous report 'Mesoscale Benchmark Demonstration, Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing', we developed a phase-field model to simulate the intra-granular gas bubble evolution in a single crystal during post-irradiation thermal annealing. In this work, we enhanced the model by incorporating thermodynamic and kinetic properties at grain boundaries, which can be obtained from atomistic simulations, to simulate fission gas bubble growth kinetics in polycrystalline UO2 fuels. The model takes into account of gas atom and vacancy diffusion, vacancy trapping and emission at defects, gas atom absorption and resolution at gas bubbles, internal pressure in gas bubbles, elastic interaction between defects and gas bubbles, and the difference of thermodynamic and kinetic properties in matrix and grain boundaries. We applied the model to simulate gas atom segregation at grain boundaries and the effect of interfacial energy and gas mobility on gas bubble morphology and growth kinetics in a bi-crystal UO2 during post-irradiation thermal annealing. The preliminary results demonstrate that the model can produce the equilibrium thermodynamic properties and the morphology of gas bubbles at

  12. Fission product transport and behavior during two postulated loss of flow transients in the air

    SciTech Connect

    Adams, J.P.; Carboneau, M.L.

    1991-01-01

    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  13. Fission product transport and behavior during two postulated loss of flow transients in the air

    SciTech Connect

    Adams, J.P.; Carboneau, M.L.

    1991-12-31

    This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10{sup {minus}5 }and 10{sup {minus}7} per reactor year for LCP15 and LPP9, respectively.

  14. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  15. 81929 - Fission-Product Separation Based on Room - Temperature Ionic Liquids

    SciTech Connect

    Robin D. Rogers

    2004-12-09

    This project has demonstrated that Sr2+ and Cs+ can be selectively extracted from aqueous solutions into ionic liquids using crown ethers and that unprecedented large distribution coefficients can be achieved for these fission products. The volume of secondary wastes can be significantly minimized with this new separation technology. Through the current EMSP funding, the solvent extraction technology based on ionic liquids has been shown to be viable and can potentially provide the most efficient separation of problematic fission products from high level wastes. The key results from the current funding period are the development of highly selective extraction process for cesium ions based on crown ethers and calixarenes, optimization of selectivities of extractants via systematic change of ionic liquids, and investigation of task-specific ionic liquids incorporating both complexant and solvent characteristics.

  16. Behavior of Cs, I, and Te in the fission product release program at ORNL

    SciTech Connect

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

    1984-01-01

    Experiments have been conducted at ORNL with highly irradiated light-water reactor (PWR and BWR) fuel rod segments to investigate fission product release in steam in the temperature range 500 to 2000/sup 0/C. Objectives were to quantify and characterize the releases under conditions postulated for LOCA) and severe accident conditions. In all, 26 experiments have been conducted - 24 with high burnup and 2 with low burnup fuels. To aid in the interpretation of fission product release, 12 implant and 18 control experiments were also conducted; the behavior of HI, I/sub 2/, Cs/sub 2/O, CsOH, Te, and TeO/sub 2/ (individually and in different combinations) was studied. This paper discusses only the observed behavior of cesium, iodine, and tellurium. Cs and I were released primarily as CsOH and CsI, and Te release was controlled by steam oxidation of Zircaloy cladding.

  17. 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.

  18. Production of Mass-Separated Fission Fragment Beams at ALTO

    SciTech Connect

    Lebois, M.; Cheikh Mhamed, M.; Curaudeau, J. M.; Ducourtieux, M.; Essabaa, S.; Franchoo, S.; Gales, S.; Guillemaud-Mueller, D.; Ibrahim, F.; Lau, C.; Lesrel, J.; Mueller, A.; Raynaud, M.; Roussiere, B.; Said, A.; Verney, D.; Vogel, C.

    2007-05-22

    Yields of neutron-rich isotopes produced by the photofission were measured at the ISOL ALTO facility. The identification was achieved by a combined measurement of {beta} and {gamma}-rays. Production rates for Xe, Kr, Sn, In and I isotopes are presented here. In parallel, empirical estimations for the yields based on the PARRNe experimental data and the results provided by a very recent FLUKA simulation are presented.

  19. METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

    DOEpatents

    Jaffey, A.H.; Seaborg, G.T.

    1958-12-23

    The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

  20. SELECTIVE SEPARATION OF URANIUM FROM THORIUM, PROTACTINIUM AND FISSION PRODUCTS BY PEROXIDE DISSOLUTION METHOD

    DOEpatents

    Seaborg, G.T.; Gofman, J.W.; Stoughton, R.W.

    1959-08-18

    A method is described for separating U/sup 233/ from thorium and fission products. The separation is effected by forming a thorium-nitric acid solution of about 3 pH, adding hydrogen peroxide to precipitate uranium and thorium peroxide, treating the peroxides with sodium hydroxide to selectively precipitate the uranium peroxide, and reacting the separated solution with nitric acid to re- precipitate the uranium peroxide.

  1. PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION

    DOEpatents

    Ellison, C.V.; Runion, T.C.

    1961-06-27

    An aqueous nitric acid solution containing uranium, plutonium, and fission product values is contacted with an organic extractant comprised of a trialkyl phosphate and an organic diluent. The relative amounts of trialkyl phosphate and uranium values are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 moles uranium per mole of trialkyl phosphate, thereby preferentially extracting uranium values into the organic extractant.

  2. Reactive transport modelling of the interaction of fission product ground contamination with alkaline and cementitious leachates

    SciTech Connect

    Kwong, S.; Small, J.

    2007-07-01

    The fission products Cs-137 and Sr-90 are amongst the most common radionuclides occurring in ground contamination at the UK civil nuclear sites. Such contamination is often associated with alkaline liquids and the mobility of these fission products may be affected by these chemical conditions. Similar geochemical effects may also result from cementitious leachate associated with building foundations and the use of grouts to remediate ground contamination. The behaviour of fission products in these scenarios is a complex interaction of hydrogeological and geochemical processes. A suite of modelling tools have been developed to investigate the behaviour of a radioactive plume containing Cs and Sr. Firstly the effects of sorption due to cementitious groundwater is modelled using PHREEQC. This chemical model is then incorporated into PHAST for the 3-D reactive solute transport modeling. Results are presented for a generic scenario including features and processes that are likely to be relevant to a number of civil UK nuclear sites. Initial results show that modelling can be a very cost-effective means to study the complex hydrogeological and geochemical processes involved. Modelling can help predict the mobility of contaminants in a range of site end point scenarios, and in assessing the consequences of decommissioning activities. (authors)

  3. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

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

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) 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 in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  4. 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 χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (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

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

    SciTech Connect

    Versey, Joshua R.

    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.

  6. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  7. Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

    SciTech Connect

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

    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 source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔGrxn°(TC))/(RTC)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn(TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2130 K and 2220 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, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

  8. ACRR (Annular Core Research Reactor) fission product release tests: ST-1 and ST-2

    SciTech Connect

    Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

    1988-01-01

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs.

  9. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    NASA Astrophysics Data System (ADS)

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.

    2005-07-01

    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 of the compressor, turbine and the electric generator (three phase AC alternator); and the heat rejection subsystem, principally the space radiator, which enables the hot gas working fluid, emanating from either the turbine or a regenerative heat exchanger, to be cooled to compressor inlet conditions. Numerical mass models for all major subsystems and components developed during the course of this work are included in this report. The power systems modeled are applicable to future interplanetary missions within the Solar System and planetary surface power plants at mission destinations, such as our Moon, Mars, the Galilean moons (Io, Europa, Ganymede, and Callisto), or Saturn's moon Titan. The detailed governing equations for the thermodynamic processes of the Brayton cycle have been derived and successfully programmed along with the heat transfer processes associated with cycle heat exchangers and the space radiator. System performance and mass results have been validated against a commercially available non-linear optimization code and also against data from existing ground based power plants.

  12. Desulfurized gas production from vertical kiln pyrolysis

    DOEpatents

    Harris, Harry A.; Jones, Jr., John B.

    1978-05-30

    A gas, formed as a product of a pyrolysis of oil shale, is passed through hot, retorted shale (containing at least partially decomposed calcium or magnesium carbonate) to essentially eliminate sulfur contaminants in the gas. Specifically, a single chambered pyrolysis vessel, having a pyrolysis zone and a retorted shale gas into the bottom of the retorted shale zone and cleaned product gas is withdrawn as hot product gas near the top of such zone.

  13. Experimental Determination of the Antineutrino Spectrum of the Fission Products of U238

    NASA Astrophysics Data System (ADS)

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

    2014-03-01

    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 β spectra were recorded with a γ-suppressing electron telescope. The obtained β 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 β spectrum of U238 was converted into the corresponding ν¯e spectrum. The final ν¯e spectrum is given in 250 keV bins in the range from 2.875 to 7.625 MeV with an energy-dependent error of 3.5% at 3 MeV, 7.6% at 6 MeV, 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.

  14. Fission product release and survivability of UN-kernel LWR TRISO fuel

    SciTech Connect

    T. M. Besmann; M. K. Ferber; H.-T. Lin; B. P. Collin

    2014-05-01

    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 fission product recoil 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 um 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 within a TRISO particle undergoing burnup. 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 computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. 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.

  15. Fission product release and survivability of UN-kernel LWR TRISO fuel

    NASA Astrophysics Data System (ADS)

    Besmann, T. M.; Ferber, M. K.; Lin, H.-T.; Collin, B. P.

    2014-05-01

    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 fission product recoil 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 μm 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 within a TRISO particle undergoing burnup. 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 computing the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers from internal pressure and thermomechanics of the layers. 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.

  16. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    SciTech Connect

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  17. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    NASA Astrophysics Data System (ADS)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  18. General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors

    Energy Science and Technology Software Center (ESTSC)

    1981-12-01

    EPRI-CINDER calculates, for any specified initial fuel (actinide) description and flux or power history, the fuel and fission-product nuclide concentrations and associated properties. Other nuclide chains can also be computed with user-supplied libraries. The EPRI-CINDER Data Library (incorporating ENDF/B-IV fission-product processed 4-group cross sections, decay constants, absorption and decay branching fractions, and effective fission yields) is used in each constant-flux time step calculation and in time step summaries of nuclide decay rates and macroscopic absorptionmore » and barns-per-fission (b/f) absorption cross sections (by neutron group). User-supplied nuclide decay energy and multigroup-spectra data libraries may be attached to permit decay heating and decay-spectra calculations. An additional 12-chain library, explicitly including 27 major fission-product neutron absorbers and 4 fictitious nuclides, may be used to accurately calculate the aggregate macroscopic absorption buildup in fission products.« less

  19. 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.

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

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    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.

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

    DOEpatents

    Norman, Eric B.; Prussin, Stanley G.

    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.

  2. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    NASA Astrophysics Data System (ADS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  3. Measurement of Short-Lived Fission-Product Yields of URANIUM-235 Using High-Resolution Gamma Spectra.

    NASA Astrophysics Data System (ADS)

    Tipnis, Sameer Vijay

    Independent yields of short-lived fission products produced by the thermal neutron induced fission of ^{235}U were determined from the measurements of high resolution gamma spectra. Comparisons were made to the recommended yield values tabulated in the ENDF/B-VI evaluated fission-product data base. Measurements of the gamma spectra were made with a high purity germanium detector (HPGe) using a NaI(Tl) annulus for Compton suppression. Use of beta-gamma coincidence reduced the random background and also allowed a precise definition of the delay time. The experiment was carried out at the 5.5 MV Van de Graaff facility at the University of Massachusetts Lowell. Rapid transfer of the fission fragments to a low background counting environment, a crucial factor in determining the yields of short-lived fission products, was enabled by a helium -jet tape transport system. The recommended yields in the evaluated data file are a combination of experimental and model-predicted values. The latter source is used since data from many short-lived fission products is still missing or poorly known. The results presented here, especially the ones for the very short-lived isotopes may be used to reduce the uncertainties associated with some of the existing values or to replace model-predicted yields. Gaussian distributions of elemental yields, based on the set of experimentally determined independent yields were examined. The feasibility of predicting unmeasured yields on the basis of charge and mass complementarity was also addressed.

  4. A model for the influence of microstructure, precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels

    NASA Astrophysics Data System (ADS)

    Rest, J.

    2004-03-01

    Irradiation-induced recrystallization appears to be a general phenomenon in that it is observed to occur in a variety of nuclear fuel types, e.g. U-xMo, UO2, and U3O8. For temperatures below that where significant thermal annealing of defects occurs, an expression is derived for the fission density at which irradiation-induced recrystallization is initiated that is athermal and weakly dependent on fission rate. The initiation of recrystallization is to be distinguished from the subsequent progression and eventual consumption of the original fuel grain. The formulation takes into account the observed microstructural evolution of the fuel, the role of precipitate pinning and fission gas bubbles, and the triggering event for recrystallization. The calculated dislocation density, fission gas bubble-size distribution, and fission density at which recrystallization first appears are compared to measured quantities.

  5. Deposition of fission and activation products after the Fukushima Dai-ichi nuclear power plant accident.

    PubMed

    Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

    2012-04-01

    The Great Eastern Japan Earthquake on March 11, 2011, damaged reactor cooling systems at Fukushima Dai-ichi nuclear power plant. The subsequent venting operation and hydrogen explosion resulted in a large radioactive nuclide emission from reactor containers into the environment. Here, we collected environmental samples such as soil, plant species, and water on April 10, 2011, in front of the power plant main gate as well as 35 km away in Iitate village, and observed gamma-rays with a Ge(Li) semiconductor detector. We observed activation products ((239)Np and (59)Fe) and fission products ((131)I, (134)Cs ((133)Cs), (137)Cs, (110m)Ag ((109)Ag), (132)Te, (132)I, (140)Ba, (140)La, (91)Sr, (91)Y, (95)Zr, and (95)Nb). (239)Np is the parent nuclide of (239)Pu; (59)Fe are presumably activation products of (58)Fe obtained by corrosion of cooling pipes. The results show that these activation and fission products, diffused within a month of the accident. PMID:22266366

  6. Progress towards the production of the 236gNp standard sources and competing fission fragment production

    NASA Astrophysics Data System (ADS)

    Larijani, C.; Pickford, O. L.; Collins, S. M.; Ivanov, P.; Jerome, S. M.; Keightley, J. D.; Pearce, A. K.; Regan, P. H.

    2015-11-01

    The isobaric distribution of fission residues produced following the bombardment of a natural uranium target with a beam of 25 MeV protons has been evaluated. Decay analysis of thirteen isobarically distinct fission residues were carried out using high-resolution γ-spectrometry at the UK National Physical Laboratory. Stoichiometric abundances were calculated via the determination of absolute activity concentrations associated with the longest-lived members of each isobaric chain. This technique was validated by computational modelling of likely sequential decay processes through an isobaric decay chain. The results were largely in agreement with previously published values for neutron bombardments on 238U at energies of 14 MeV. Higher yields of products with mass numbers A~110-130 were found, consistent with the increasing yield of these radionuclides as the bombarding energy is increased.

  7. Measurements of fission product yield in the neutron-induced fission of 238U with average energies of 9.35 MeV and 12.52 MeV

    NASA Astrophysics Data System (ADS)

    Mukerji, Sadhana; Krishnani, Pritam Das; Shivashankar, Byrapura Siddaramaiah; Mulik, Vikas Kaluram; Suryanarayana, Saraswatula Venkat; Naik, Haladhara; Goswami, Ashok

    2014-07-01

    The yields of various fission products in the neutron-induced fission of 238U with the flux-weightedaveraged neutron energies of 9.35 MeV and 12.52 MeV were determined by using an off-line gammaray spectroscopic technique. The neutrons were generated using the 7Li(p, n) reaction at Bhabha Atomic Research Centre-Tata Institute of Fundamental Research Pelletron facility, Mumbai. The gamma- ray activities of the fission products were counted in a highly-shielded HPGe detector over a period of several weeks to identify the decaying fission products. At both the neutron energies, the fission-yield values are reported for twelve fission product. The results obtained from the present work have been compared with the similar data for mono-energetic neutrons of comparable energy from the literature and are found to be in good agreement. The peak-to-valley (P/V) ratios were calculated from the fission-yield data and were found to decreases for neutron energy from 9.35 to 12.52 MeV, which indicates the role of excitation energy. The effect of the nuclear structure on the fission product-yield is discussed.

  8. Fission-product data analysis from actinide samples exposed in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Murphy, B.D.; Dickens, J.K.; Walker, R.L.; Newton, T.D.

    1994-12-31

    Since 1979 a cooperative agreement has been in effect between the United States and the United Kingdom to investigate the irradiation of various actinide species placed in the core of the Dounreay Prototype Fast Reactor (PFR). The irradiated species were isotopes of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium. A set of actinide samples (mg quantities) was exposed to about 490 effective full power days (EFPD) of reactor operations. The fission-product results are reported here. The actinide results will be report elsewhere.

  9. Design and Expected Performance of the AGR-1 Fission Product Monitoring System (FPMS)

    SciTech Connect

    John K. Hartwell; Dawn M. Scates

    2005-09-01

    The effluent from each test capsule of the AGR-1 experiment will be monitored by a detector system consisting of a gamma-ray spectrometer and a gross radiation detector. This collection of radiation measurement systems will be known as the AGR-1 Fission Product Monitoring System (FPMS). Proper design and functioning of the FPMS is critical to the success of the AGR-1 fuel test experiment.This document describes the AGR-1 FPMS and presents calculations indicating that this design will meet the pertinent test requirements.

  10. 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.

  11. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    SciTech Connect

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-08-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ``engineered safety features,`` which, along with the use of high temperature capable materials further enhance its safety characteristics.

  12. Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

    SciTech Connect

    Wichner, R.P.; Weber, C.F.; Hodge, S.A.; Beahm, E.C.; Wright, A.L.

    1984-01-01

    This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor.

  13. Delayed-neutron branching ratios of precursors in the fission product region

    SciTech Connect

    Rudstam, G.; Aleklett, K.; Sihver, L. )

    1993-01-01

    Delayed-neutron branching ratios in the fission product region have been tabulated, and average values have been determined. In order to provide data complementary to published values an experiment covering the mass range 79-150 has been carried out at the OSIRIS isotope-separator on-line facility at Studsvik. This experiment has resulted in branching ratios for some precursors ([sup 84]Ge, [sup 133]Sn, and [sup 150]La) for which such data have not been reported before. In several other cases the new results are accurate than older determinations. 19 refs., 2 figs., 1 tab.

  14. Properties of the platinoid fission products during vitrification of high-level radioactive waste

    NASA Astrophysics Data System (ADS)

    Gong, W.; Lutze, W.; Perez-Cardenas, F.; Matlack, K. S.; Pegg, I. L.

    2006-05-01

    Platinoid fission products present in high-level nuclear wastes present particular challenges to their treatment by vitrification. The platinoid metals Ru, Rh, Pd, and their compounds are sparingly soluble in borosilicate glass melts. During glass melting under oxidizing conditions, the platinoids form small crystals of highly dense solid intermetallic phases and oxides. Under reducing conditions, the platinoids form only intermetallic phases. A fraction of these crystals settles to the bottom of the melting furnace, forming an immobile sludge. The fraction settling reported in the literature is highly variable. In the present work, the fraction settling was found to be >90% under reducing conditions but only 10 to 20% under oxidizing conditions. The thickness of the sludge layer depends on the volume fraction of platinoid crystals in the sludge, which is poorly known (typically ~0.06 under oxidizing conditions). Since the electrical conductivity of the sludge can be >10X that of the melt, in joule-heated melters the presence of such a layer can lead to diversion of the electric current, thereby compromising melter operability. The time to failure by this mechanism is clearly of practical importance. A variety of data are required in order to estimate the time to failure due to this mechanism and such data must be obtained under conditions representative of those in a full-size melting furnace. We have acquired such data using a melting furnace installed in our laboratory. This furnace is a one-third scale prototype of the system to be used for the vitrification of defense HLW at Hanford, WA. In the present work, simulated Hanford HLW material was combined with glass formers to produce a melter feed slurry that was then spiked with the platinoids. Over one thousand chemical and optical analyses were performed on hundreds of samples taken from the feed, various locations inside the furnace, the glass melt during pouring, the solid glass, and various locations along

  15. Towards the inclusion of open fabrication porosity in a fission gas release model

    NASA Astrophysics Data System (ADS)

    Claisse, Antoine; Van Uffelen, Paul

    2015-11-01

    A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature.

  16. Monte Carlo Models for the Production of beta-delayed Gamma Rays Following Fission of Special Nuclear Materials

    SciTech Connect

    Pruet, J; Prussin, S; Descalle, M; Hall, J

    2004-02-03

    A Monte Carlo method for the estimation of {beta}-delayed {gamma}-ray spectra following fission is described that can accommodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each decay mode. Though computationally intensive, the method can provide a detailed estimate of the spectrum that would be recorded by an arbitrary spectrometer, and can prove useful in assessing the quality of evaluated data libraries, for identifying gaps in these libraries, etc. The method is illustrated by a first comparison of calculated and experimental spectra from decay of short-lived fission products following the reactions {sup 235}U(n{sub th}, f) and {sup 239}Pu(n{sub th}, f). For general purpose transport calculations, where detailed consideration of the large number of individual {gamma}-ray transitions in a spectrum may be unnecessary, it is shown that an accurate and simple parameterization of a {gamma}-ray source function can be obtained. These parametrizations should provide high-quality average spectral distributions that should prove useful in calculations describing photons escaping from thick attenuating media.

  17. Krypton and xenon in Apollo 14 samples - Fission and neutron capture effects in gas-rich samples

    NASA Technical Reports Server (NTRS)

    Drozd, R.; Hohenberg, C.; Morgan, C.

    1975-01-01

    Gas-rich Apollo 14 breccias and trench soil are examined for fission xenon from the decay of the extinct isotopes Pu-244 and I-129, and some samples have been found to have an excess fission component which apparently was incorporated after decay elsewhere and was not produced by in situ decay. Two samples have excess Xe-129 resulting from the decay of I-129. The excess is correlated at low temperatures with excess Xe-128 resulting from neutron capture on I-127. This neutron capture effect is accompanied by related low-temperature excesses of Kr-80 and Kr-82 from neutron capture on the bromine isotopes. Surface correlated concentrations of iodine and bromine are calculated from the neutron capture excesses.

  18. 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.

  19. Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2016-05-01

    In this study, the fission product precipitates at silicon carbide grain boundaries from an irradiated TRISO particle were identified and correlated with the associated grain boundary characteristics. Precession electron diffraction in the transmission electron microscope provided the crystallographic information needed to identify grain boundary misorientation and boundary type (i.e., low angle, random high angle or coincident site lattice (CSL)-related). The silicon carbide layer was found to be composed mainly of twin boundaries and small fractions of random high angle and low angle grain boundaries. Most fission products were found at random, high-angle grain boundaries, with small fractions at low-angle and CSL-related grain boundaries. Palladium (Pd) was found at all types of grain boundaries while Pd-uranium and Pd-silver precipitates were only associated with CSL-related and random, high-angle grain boundaries. Precipitates containing only Ag were found only at random, high-angle grain boundaries, but not at low angle or CSL-related grain boundaries.

  20. METHOD FOR SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY SOLVENT EXTRACTION

    DOEpatents

    Seaborg, G.T.; Blaedel, W.J.; Walling, M.T. Jr.

    1960-08-23

    A process is given for separating from each other uranium, plutonium, and fission products in an aqueous nitric acid solution by the so-called Redox process. The plutonium is first oxidized to the hexavalent state, e.g., with a water-soluble dichromate or sodium bismuthate, preferably together with a holding oxidant such as potassium bromate. potassium permanganate, or an excess of the oxidizing agent. The solution is then contacted with a water-immiscible organic solvent, preferably hexone. whereby uranium and plutonium are extracted while the fission products remain in the aqueous solution. The separated organic phase is then contacted with an aqueous solution of a reducing agent, with or without a holding reductant (e.g., with a ferrous salt plus hydrazine or with ferrous sulfamate), whereby plutonium is reduced to the trivalent state and back- extracted into the aqueous solution. The uranium may finally be back-extracted from the organic solvent (e.g., with a 0.1 N nitric acid).

  1. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    NASA Astrophysics Data System (ADS)

    Shcherbina, Natalia; Kivel, Niko; Günther-Leopold, Ines

    2013-06-01

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H2/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology.

  2. Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

    SciTech Connect

    Martin, R.C.

    1990-01-01

    A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of {sup 137}Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for {sup 85}Kr, {sup 90}Sr, and {sup 110m}Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models in fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. An oxygen-enhanced SiC decomposition mechanism is also discussed. 12 refs., 11 figs., 2 tabs.

  3. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  4. The chemical state of fission products in oxide fuels at different stages of the nuclear fuel cycle

    SciTech Connect

    Kleykamp, H.

    1988-03-01

    A survey of work at the Kernforschungszentrum Karlsruhe is presented on the chemical state of selected fission products that are relevant in the fuel cycle of light water reactor (LWR) and fast breeder reactor fuels. The influence of fuel type and irradiation progress on the composition of the Mo-Tc-Ru-Rh-Pd fission product alloys precipitated in the oxide matrix is examined using the respective multicomponent phase diagrams. The kinetics of dissolution of these phases in nitric acid at the reprocessing stage is discussed. Composition and structure of the residues, and the reprecipitation phenomena from highly active waste (HAW), are elucidated. A second metamorphosis of the fission products is recognized during the vitrification process. The formation of Ru(Rh) oxide and Pd(Rh, U, Te) alloys in simulated vitrified HAW concentrate and in HAW concentrate from the reprocessing of irradiated LWR fuels in interpreted on the basis of heterogeneous equilibria.

  5. The DART dispersion analysis research tool: A mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. User`s guide for mainframe, workstation, and personal computer applications

    SciTech Connect

    Rest, J.

    1995-08-01

    This report describes the primary physical models that form the basis of the DART mechanistic computer model for calculating fission-product-induced swelling of aluminum dispersion fuels; the calculated results are compared with test data. In addition, DART calculates irradiation-induced changes in the thermal conductivity of the dispersion fuel, as well as fuel restructuring due to aluminum fuel reaction, amorphization, and recrystallization. Input instructions for execution on mainframe, workstation, and personal computers are provided, as is a description of DART output. The theory of fission gas behavior and its effect on fuel swelling is discussed. The behavior of these fission products in both crystalline and amorphous fuel and in the presence of irradiation-induced recrystallization and crystalline-to-amorphous-phase change phenomena is presented, as are models for these irradiation-induced processes.

  6. Review of ENDF/B-VI Fission-Product Cross Section

    SciTech Connect

    Wright, R.Q.

    1999-01-01

    the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature.

  7. Fission-fragment excited xenon/rare gas mixtures. I. Laser parameters of the 1. 73 [mu]m xenon transition

    SciTech Connect

    Hebner, G.A.; Hays, G.N. )

    1993-04-15

    Laser parameters for the 1.73 [mu]m (5[ital d][3/2][sub 1][minus]6[ital p][5/2][sub 2]) xenon transition in fission-fragment excited Ar/Xe, He/Ar/Xe, Ne/Ar/Xe, and He/Ne/Ar/Xe gas mixtures are presented. Using a cw F center laser, time resolved small signal gain was probed as a function of total pressure, xenon concentration, pump power, He/Ne/Ar buffer ratio and impurity concentration. Small signal gains of up to 2%/cm were observed for pump rates of 30 W/cm[sup 3]. Addition of helium and/or neon to the argon buffer increased the width of the time resolved laser gain pulse and reduced the absorption observed under some experimental conditions. Experimentally determined gain scaling laws for several gas mixtures are presented. The measured small signal gain was coupled with the results of laser cavity measurements to calculate the saturation intensity for several gas mixtures. The addition of helium or neon increases the saturation intensity for several gas mixtures. Laser cavity measurements as well as the gain [times] saturation intensity product indicate that the 1.73 [mu]m power efficiency is approximately 2% for several gas mixtures.

  8. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... production, by final product sold, of oil, gas, and other products. Disclosure shall be made by geographical... conversion to synthetic oil or gas, the product's production, transfer prices, and production costs should be... expressed in common units of production with oil, gas, and other products converted to a common unit...

  9. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... production, by final product sold, of oil, gas, and other products. Disclosure shall be made by geographical... conversion to synthetic oil or gas, the product's production, transfer prices, and production costs should be... expressed in common units of production with oil, gas, and other products converted to a common unit...

  10. 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.

  11. 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.

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

    SciTech Connect

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

    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.

  13. New antineutrino energy spectra predictions from the summation of beta decay branches of the fission products.

    PubMed

    Fallot, M; Cormon, S; Estienne, M; Algora, A; Bui, V M; Cucoanes, A; Elnimr, M; Giot, L; Jordan, D; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Taín, J L; Yermia, F; Zakari-Issoufou, A-A

    2012-11-16

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the (102;104;105;106;107)Tc, (105)Mo, and (101)Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes (235,238)U and (239,241)Pu. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of (239)Pu, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of (235)U, (239,241)Pu, and, in particular, (238)U for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra. PMID:23215477

  14. New Antineutrino Energy Spectra Predictions from the Summation of Beta Decay Branches of the Fission Products

    NASA Astrophysics Data System (ADS)

    Fallot, M.; Cormon, S.; Estienne, M.; Algora, A.; Bui, V. M.; Cucoanes, A.; Elnimr, M.; Giot, L.; Jordan, D.; Martino, J.; Onillon, A.; Porta, A.; Pronost, G.; Remoto, A.; Taín, J. L.; Yermia, F.; Zakari-Issoufou, A.-A.

    2012-11-01

    In this Letter, we study the impact of the inclusion of the recently measured beta decay properties of the Tc102;104;105;106;107, Mo105, and Nb101 nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes U235,238 and Pu239,241. These actinides are the main contributors to the fission processes in pressurized water reactors. The beta feeding probabilities of the above-mentioned Tc, Mo, and Nb isotopes have been found to play a major role in the γ component of the decay heat of Pu239, solving a large part of the γ discrepancy in the 4-3000 s range. They have been measured by using the total absorption technique, insensitive to the pandemonium effect. The calculations are performed by using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of U235, Pu239,241, and, in particular, U238 for which no measurement has been published yet. We conclude that new total absorption technique measurements are mandatory to improve the reliability of the predicted spectra.

  15. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... production, by final product sold, of oil, gas, and other products. Disclosure shall be made by geographical... conversion to synthetic oil or gas, the product's production, transfer prices, and production costs should be... (Extractive Activities—Oil and Gas Topic). Instruction 5 to Item 1204: The average production cost,...

  16. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... production, by final product sold, of oil, gas, and other products. Disclosure shall be made by geographical... conversion to synthetic oil or gas, the product's production, transfer prices, and production costs should be... (Extractive Activities—Oil and Gas Topic). Instruction 5 to Item 1204: The average production cost,...

  17. Laboratory-Scale Bismuth Phosphate Extraction Process Simulation To Track Fate of Fission Products

    SciTech Connect

    Serne, R. JEFFREY; Lindberg, Michael J.; Jones, Thomas E.; Schaef, Herbert T.; Krupka, Kenneth M.

    2007-02-28

    Recent field investigation that collected and characterized vadose zone sediments from beneath inactive liquid disposal facilities at the Hanford 200 Areas show lower than expected concentrations of a long-term risk driver, Tc-99. Therefore laboratory studies were performed to re-create one of the three processes that were used to separate the plutonium from spent fuel and that created most of the wastes disposed or currently stored in tanks at Hanford. The laboratory simulations were used to compare with current estimates based mainly on flow sheet estimates and spotty historical data. Three simulations of the bismuth phosphate precipitation process show that less that 1% of the Tc-99, Cs-135/137, Sr-90, I-129 carry down with the Pu product and thus these isotopes should have remained within the metals waste streams that after neutralization were sent to single shell tanks. Conversely, these isotopes should not be expected to be found in the first and subsequent cycle waste streams that went to cribs. Measurable quantities (~20 to 30%) of the lanthanides, yttrium, and trivalent actinides (Am and Cm) do precipitate with the Pu product, which is higher than the 10% estimate made for current inventory projections. Surprisingly, Se (added as selenate form) also shows about 10% association with the Pu/bismuth phosphate solids. We speculate that the incorporation of some Se into the bismuth phosphate precipitate is caused by selenate substitution into crystal lattice sites for the phosphate. The bulk of the U daughter product Th-234 and Np-237 daughter product Pa-233 also associate with the solids. We suspect that the Pa daughter products of U (Pa-234 and Pa-231) would also co-precipitate with the bismuth phosphate induced solids. No more than 1 % of the Sr-90 and Sb-125 should carry down with the Pu product that ultimately was purified. Thus the current scheme used to estimate where fission products end up being disposed overestimates by one order of magnitude the

  18. 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.

  19. Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

    SciTech Connect

    Cols, H.; Cristini, P.; Marques, R.

    1997-08-01

    The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing high enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.

  20. Fission Product Release from Molten U/Al Alloy Fuel: A Vapor Transpiration Model

    SciTech Connect

    Whitkop, P.G.

    2001-06-26

    This report describes the application of a vapor transportation model to fission product release data obtained for uranium/aluminum alloy fuel during early Oak Ridge fuel melt experiments. The Oak Ridge data validates the vapor transpiration model and suggests that iodine and cesium are released from the molten fuel surface in elemental form while tellurium and ruthenium are released as oxides. Cesium iodide is postulated to form in the vapor phase outside of the fuel matrix. Kinetic data indicates that cesium iodide can form from Cs atoms and diatomic iodine in the vapor phase. Temperatures lower than those capable of melting fuel are necessary in order to maintain a sufficient I2 concentration. At temperatures near the fuel melting point, cesium can react with iodine atoms to form CsI only on solid surfaces such as aerosols.

  1. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    SciTech Connect

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  2. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Rasmussen, G.; Konings, R. J. M.

    2015-10-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the evaluation of the source term and consequently the risk associated with release from spent fuel sabotage or accidents. Different simulated spent fuels were tested with burn-up up to 8 at. %. The results from the aerosol characterisation were compared with studies of the vaporization process by Knudsen Effusion Mass Spectrometry and thermochemical equilibrium calculations. These studies permit an understanding of the aerosol gaseous precursors and the gaseous reactions taking place during the aerosol formation process.

  3. 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.

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

    SciTech Connect

    Conner, C.; Chamberlain, D.B.; Chen, L.

    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.

  5. 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.

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

    SciTech Connect

    Wichner, R.P. )

    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.

  7. 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 1072°C; a time average, peak temperature of 1182°C 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.

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

    SciTech Connect

    Liu, Xiang-yand; Uberuaga, Blas P; Nerikar, Pankaj; Sickafus, Kurt E; Stanek, Chris R

    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.

  9. Atomic scale mobility of the volatile fission products Xe, Kr and I in cubic SiC.

    PubMed

    Cooper, M W D; Kelly, S; Bertolus, M

    2016-06-22

    The migration barriers for the vacancy-assisted migration of fission products in 3C-SiC are reported and analysed in the context of the five frequency model, which enables one to calculate an effective diffusion coefficient from elementary mechanisms. Calculations were carried out using the nudged elastic band method (NEB) with interatomic forces determined from density functional theory (DFT). Justification for treating vacancy-assisted fission product migration as limited to the FCC carbon sublattice is based on the stability of carbon vacancies, unfavourable silicon vacancy formation and the accommodation of fission products on the carbon sublattice. Results show that for most Fermi levels within the band gap the activation energy for I exceeds that of Xe which exceeds that of Kr. Results also indicate that activation energies are higher near the conduction edge, thus, implying that enhanced fission product retention can be achieved through n-type doping of 3C-SiC, which limits the availability of the migration mediating carbon vacancies. PMID:27282287

  10. Interaction of fission products and SiC in TRISO fuel particles: a limiting HTGR design parameter

    SciTech Connect

    Stansfield, O.M.; Homan, F.J.; Simon, W.A.; Turner, R.F.

    1983-09-01

    The fuel particle system for the steam cycle cogeneration HTGR being developed in the US consists of 20% enriched UC/sub 0/./sub 3/O/sub 1/./sub 7/ and ThO/sub 2/ kernels with TRISO coatings. The reaction of fission products with the SiC coating is the limiting thermochemical coating failure mechanism affecting performance. The attack of the SiC by palladium (Pd) is considered the controlling reaction with systems of either oxide or carbide fuels. The lanthanides, such as cerium, neodymium, and praseodymium, also attack SiC in carbide fuel particles. In reactor design, the time-temperature relationships at local points in the core are used to calculate the depth of SiC-Pd reaction. The depth of penetration into the SiC during service varies with core power density, power distribution, outlet gas temperature, and fuel residence time. These parameters are adjusted in specifying the core design to avoid SiC coating failure.

  11. How EIA Estimates Natural Gas Production

    EIA Publications

    2004-01-01

    The Energy Information Administration (EIA) publishes estimates monthly and annually of the production of natural gas in the United States. The estimates are based on data EIA collects from gas producing states and data collected by the U. S. Minerals Management Service (MMS) in the Department of Interior. The states and MMS collect this information from producers of natural gas for various reasons, most often for revenue purposes. Because the information is not sufficiently complete or timely for inclusion in EIA's Natural Gas Monthly (NGM), EIA has developed estimation methodologies to generate monthly production estimates that are described in this document.

  12. Substrate preference of citrus naringenin rhamnosyltransferases and their application to flavonoid glycoside production in fission yeast.

    PubMed

    Ohashi, Takao; Hasegawa, Yuka; Misaki, Ryo; Fujiyama, Kazuhito

    2016-01-01

    Flavonoids, which comprise a large family of secondary plant metabolites, have received increased attention in recent years due to their wide range of features beneficial to human health. One of the most abundant flavonoid skeletons in citrus species is the flavanone naringenin, which is accumulated as glycosides containing terminal rhamnose (Rha) after serial glycosylation steps. The linkage type of Rha residues is a determining factor in the bitterness of the citrus fruit. Such Rha residues are attached by either an α1,2- or an α1,6-rhamnosyltransferase (1,2RhaT or 1,6RhaT). Although the genes encoding these RhaTs from pummelo (Citrus maxima) and orange (Citrus sinensis) have been functionally characterized, the details of the biochemical characterization, including the substrate preference, remain elusive due to the lack of availability of the UDP-Rha required as substrate. In this study, an efficient UDP-Rha in vivo production system using the engineered fission yeast expressing Arabidopsis thaliana rhamnose synthase 2 (AtRHM2) gene was constructed. The in vitro RhaT assay using the constructed UDP-Rha revealed that recombinant RhaT proteins (Cm1,2RhaT; Cs1,6RhaT; or Cm1,6RhaT), which were heterologously produced in fission yeast, catalyzed the rhamnosyl transfer to naringenin-7-O-glucoside as an acceptor. The substrate preference analysis showed that Cm1,2RhaT had glycosyl transfer activity toward UDP-xylose as well as UDP-Rha. On the other hand, Cs1,6RhaT and Cm1,6RhaT showed rhamnosyltransfer activity toward quercetin-3-O-glucoside in addition to naringenin-7-O-glucoside, indicating weak specificity toward acceptor substrates. Finally, naringin and narirutin from naringenin-7-O-glucoside were produced using the engineered fission yeast expressing the AtRHM2 and the Cm1,2RhaT or the Cs1,6RhaT genes as a whole-cell-biocatalyst. PMID:26433966

  13. On the role of energy separated in fission process, excitation energy and reaction channels effects in the isomeric ratios of fission product 135Xe in photofission of actinide elements

    NASA Astrophysics Data System (ADS)

    Thiep, Tran Duc; An, Truong Thi; Cuong, Phan Viet; Vinh, Nguyen The; Mishinski, G. V.; Zhemenik, V. I.

    2016-07-01

    In this work we present the isomeric ratio of fission product 135Xe in the photo-fission of actinide elements 232Th, 233U and 237Np induced by end-point bremsstrahlung energies of 13.5, 23.5 and 25.0 MeV which were determined by the method of inert gaseous flow. The data were analyzed, discussed and compared with the similar data from literature to examine the role of energy separated in fission process, excitation energy and reaction channels effects.

  14. Partitioning of fission products from irradiated nitride fuel using inductive vaporization

    SciTech Connect

    Shcherbina, N.; Kulik, D.A.; Kivel, N.; Potthast, H.D.; Guenther-Leopold, I.

    2013-07-01

    Irradiated nitride fuel (Pu{sub 0.3}Zr{sub 0.7})N fabricated at PSI in frame of the CONFIRM project and having a burn-up of 10.4 % FIMA (Fission per Initial Metal Atom) has been investigated by means of inductive vaporization. The study of thermal stability and release behavior of Pu, Am, Zr and fission products (FPs) was performed in a wide temperature range (up to 2300 C. degrees) and on different redox conditions. On-line monitoring by ICP-MS detected low nitride stability and significant loss of Pu and Am at T>1900 C. degrees during annealing under inert atmosphere (Ar). The oxidative pre-treatment of nitride fuel on air at 1000 C. degrees resulted in strong retention of Pu and Am in the solid, as well as of most FPs. Thermodynamic modelling of elemental speciation using GEM-Selektor v.3 code (Gibbs Energy Minimization Selektor), supported by a comprehensive literature review on thermodynamics of actinides and FPs, revealed a number of binary compounds of Cs, Mo, Te, Sr and Ba to occur in the solid. Speciation of some FPs in the fuel is discussed and compared to earlier results of electron probe microanalysis (EPMA). Predominant vapor species predicted by GEM-Selektor calculations were Pu(g), Am(g) and N{sub 2}. Nitrogen can be completely released from the fuel after complete oxidation at 1000 C. degrees. With regard to the irradiated nitride reprocessing technology, this result can have an important practical application as an alternative way for {sup 15}N recovery. (authors)

  15. 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

  16. 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.

  17. Fission Xenon on Mars

    NASA Technical Reports Server (NTRS)

    Mathew, K. J.; Marti, K.; Marty, B.

    2002-01-01

    Fission Xe components due to Pu-244 decay in the early history of Mars have been identified in nakhlites; as in the case of ALH84001 and Chassigny the fission gas was assimilated into indigenous solar-type Xe. Additional information is contained in the original extended abstract.

  18. Cyclotron production of molybdenum-99 via proton-induced uranium-238 fission

    SciTech Connect

    Lagunas-Solar, M.C.; Zeng, N.X.; Mirshad, I.; Castaneda, C.M.

    1996-12-31

    Technetium-99m ({sup 99m}Tc; 6.02 h) is the most widely used radioisotope in nuclear medicine worldwide. It is currently supplied from elutions of a {sup 99}Mo {r_arrow} {sup 99m}Tc generator and used for regional distribution or locally for institutional use. The parent {sup 99}Mo (66.02 h) radioactivities are being produced commercially in reactors using the {sup 235}U(n, fission){sup 99}Mo (preferred method) or the {sup 98}Mo(n,{gamma}){sup 99}Mo (less desirable) methods. The production of {sup 99}Mo is based on the operation of a small number of nuclear reactors, most of which have reached decommissioning age. Two new reactors to be dedicated to radioisotope production are now being planned in Canada. Accelerator-based methods for producing {sup 99}Mo and/or {sup 99m}Tc would then provide new alternatives to the current reliance on reactor-based technologies.

  19. 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.

  20. Myanmar production meets first-gas targets

    SciTech Connect

    Lepage, A.

    1998-09-07

    Despite scheduling complications caused by annual monsoons, the Yadana project to bring offshore Myanmar gas ashore and into neighboring Thailand has met it first-gas target of July 1, 1998. The Yadana field is a dry-gas reservoir in the reef upper Birman limestone formation t 1,260 m and a pressure of 174 bara (approximately 2,500 psi). It extends nearly 7 km (west to east) and 10 km (south to north). The water-saturated reservoir gas contains mostly methane mixed with CO{sub 2} and N{sub 2}. No production of condensate is anticipated. The Yadana field contains certified gas reserves of 5.7 tcf, calculated on the basis of 2D and 3D seismic data-acquisition campaigns and of seven appraisal wells. The paper discusses early interest, development sequences, offshore platforms, the gas-export pipeline, safety, environmental steps, and schedule constraints.

  1. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    NASA Astrophysics Data System (ADS)

    Dubourg, R.; Barrachin, M.; Ducher, R.; Gavillet, D.; De Bremaecker, A.

    2014-10-01

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  2. 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.

    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.

  3. Hot beta particles in the lung: Results from dogs exposed to fission product radionuclides

    SciTech Connect

    Hahn, F.F.; Griffith, W.C.; Hobbs, C.H.

    1995-12-01

    The Chernobyl nuclear reactor accident resulted in the release of uranium dioxide fuel and fission product radionuclides into the environment with the fallout of respirable, highly radioactive particles that have been termed {open_quotes}hot beta particles.{close_quotes} There is concern that these hot beta particles (containing an average of 150-20,000 Bq/particle), when inhaled and deposited in the lung, may present an extraordinary hazard for the induction of lung cancer. We reviewed data from a group of studies in dogs exposed to different quantities of beta-emitting radionuclides with varied physical half-lives to determine if those that inhaled hot beta particles were at unusual risk for lung cancer. This analysis indicates that the average dose to the lung is adequate to predict biologic effects of lung cancer for inhaled beta-emitting radionuclides in the range of 5-50 Gy to the lung and with particle activities in the range of 0.10-50 Bq/particle.

  4. High-Resolution Compton-Suppressed CZT Detector for Fission Products Identification

    SciTech Connect

    R. Aryaeinejd; J. K. Hartwell; Wade W. Scates

    2004-10-01

    Room temperature semiconductor CdZnTe (CZT) detectors are currently limited to total detector volumes of 1-2 cm3, which is dictated by the poor charge transport characteristics. Because of this size limitation one of the problems in accurately determining isotope identification is the enormous background from the Compton scattering events. Eliminating this background will not only increase the sensitivity and accuracy of measurements but also help us to resolve peaks buried under the background and peaks in close vicinity of others. We are currently developing a fission products detection system based on the Compton-suppressed CZT detector. In this application, the detection system is required to operate in high radiation fields. Therefore, a small 10x10x5 mm3 CZT detector is placed inside the center of a well-shielded 3" in diameter by 3" long Nal detector. So far we have been able to successfully reduce the Compton background by a factor of 5.4 for a 137Cs spectrum. This reduction of background will definitely enhance the quality of the gamma-ray spectrum in the information-rich energy range below 1 MeV, which consequently increases the detection sensitivity. In this work, we will discuss the performance of this detection system as well as its applications.

  5. Separation of the rare-earth fission product poisons from spent nuclear fuel

    DOEpatents

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  6. Electrochemical separation of actinides and fission products in molten salt electrolyte

    NASA Astrophysics Data System (ADS)

    Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J.

    1995-09-01

    Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation of the metal chlorides are grouped advantageously such that the actinides can be deposited on a solid cathode with the majority of the rare earths remaining in the electrolyte. Thus, the actinides are recycled for further transmutation. Rockwell and its partners have measured the thermodynamic properties of the metal chlorides of interest (rare earths and actinides) and demonstrated separation of actinides from rare earths in laboratory studies. A model is being developed to predict the performance of a commercial electrochemical cell for separations starting with PUREX compositions. This model predicts excellent separation of plutonium and other actinides from the rare earths in metal-salt systems.

  7. 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.

  8. Study on collection efficiency of fission products by spray: Experimental device and modelling

    SciTech Connect

    Ducret, D.; Roblot, D.; Vendel, J.; Billarand, Y.

    1997-08-01

    Consequences of an hypothetical overheating reactor accident in nuclear power plants can be limited by spraying cold water drops into containment building. The spray reduces the pressure and the temperature levels by condensation of steam and leads to the washout of fission products (aerosols and gaseous iodine). The present study includes a large program devoted to the evaluation of realistic washout rates. An experimental device (named CARAIDAS) was designed and built in order to determine the collection efficiency of aerosols and iodine absorption by drops with representative conditions of post-accident atmosphere. This experimental device is presented in the paper and more particularly: (1) the experimental enclosure in which representative thermodynamic conditions can be achieved, (2) the monosized drops generator, the drops diameter measurement and the drops collector, (3) the cesium iodide aerosols generator and the aerosols measurements. Modelling of steam condensation on drops aerosols collection and iodine absorption are described. First experimental and code results on drops and aerosols behaviour are compared. 8 refs., 18 figs.

  9. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    NASA Astrophysics Data System (ADS)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

  10. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  11. EXTRACTION METHOD FOR SEPARATING URANIUM, PLUTONIUM, AND FISSION PRODUCTS FROM COMPOSITIONS CONTAINING SAME

    DOEpatents

    Seaborg, G.T.

    1957-10-29

    Methods for separating plutonium from the fission products present in masses of neutron irradiated uranium are reported. The neutron irradiated uranium is first dissolved in an aqueous solution of nitric acid. The plutonium in this solution is present as plutonous nitrate. The aqueous solution is then agitated with an organic solvent, which is not miscible with water, such as diethyl ether. The ether extracts 90% of the uraryl nitrate leaving, substantially all of the plutonium in the aqueous phase. The aqueous solution of plutonous nitrate is then oxidized to the hexavalent state, and agitated with diethyl ether again. In the ether phase there is then obtained 90% of plutonium as a solution of plutonyl nitrate. The ether solution of plutonyl nitrate is then agitated with water containing a reducing agent such as sulfur dioxide, and the plutonium dissolves in the water and is reduced to the plutonous state. The uranyl nitrate remains in the ether. The plutonous nitrate in the water may be recovered by precipitation.

  12. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    DOE PAGESBeta

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

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

    SciTech Connect

    MacMullin, S.; Giovanetti, G. K.; Green, M. P.; Henning, R.; Holmes, R.; Vorren, K.

    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).

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

    PubMed

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

    2012-10-01

    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 (131)I and (137)Cs were measured with maximum activity concentrations of 4.2 ± 0.6 mBq/m(3) and 0.42 ± 0.07 mBq/m(3) respectively. Additional activity from (131,132)I, (134,136,137)Cs and (132)Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF). PMID:22348994

  15. Partition of actinides and fission products between metal and molten salt phases: Theory, measurement, and application to IFR pyroprocess development

    SciTech Connect

    Ackerman, J.P.; Johnson, T.R.

    1993-10-01

    The chemical basis of Integral Fast Reactor fuel reprocessing (pyroprocessing) is partition of fuel, cladding, and fission product elements between molten LiCl-KCl and either a solid metal phase or a liquid cadmium phase. The partition reactions are described herein, and the thermodynamic basis for predicting distributions of actinides and fission products in the pyroprocess is discussed. The critical role of metal-phase activity coefficients, especially those of rare earth and the transuranic elements, is described. Measured separation factors, which are analogous to equilibrium constants but which involve concentrations rather than activities, are presented. The uses of thermodynamic calculations in process development are described, as are computer codes developed for calculating material flows and phase compositions in pyroprocessing.

  16. Experimental investigations on the chemical state of solid fission-product elements in U3Si2

    NASA Astrophysics Data System (ADS)

    Ugajin, M.; Itoh, A.

    1994-10-01

    The uranium silicide U3Si2 has a congruent melting point of 1665 C and possesses higher uranium density (11.3 g U/cc) and higher thermal conductivity than the uranium dioxide currently used in light water reactors. U3Si2 is in use as a research reactor fuel (US Nuclear Regulatory Commission, NUREG-1313, July, 1988), representing a potentiality for power reactor fuel. A first attempt is made in this study to predict the chemical state of the solid fission-product elements comprising zirconium, molybdenum, rare earth elements, alkaline earth metals and elements of the platinum group. Ternary phase equilibria in the U-Mo-Si and U-Ru-Si systems are also investigated to supplement the fission product chemistry in U3Si2.

  17. Deep Atomic Binding (DAB) Approach in Interpretation of Fission Products Behavior in Human Body, and Health Consequences

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    According to models used to predict health effects of fission products enter the human body, a large number of fatalities, malignancies, thyroid cancer, born (genetic) defects,...etc.. But the actual data after Chernobyl and TMI accidents, and nuclear detonations in USA and Marshal Islands, were not consistent with these models. According to DAB, these data could be interpreted, and conflicts between former models predictions and actual field data explained. (author)

  18. Aqueous Biphasic Systems Based on Salting-Out Polyethylene Glycol or Ionic Solutions: Strategies for Actinide or Fission Product Separations

    SciTech Connect

    Rogers, Robin D.; Gutowski, Keith E.; Griffin, Scott T.; Holbrey, John D.

    2004-03-29

    Aqueous biphasic systems can be formed by salting-out (with kosmotropic, waterstructuring salts) water soluble polymers (e.g., polyethylene glycol) or aqueous solutions of a wide range of hydrophilic ionic liquids based on imidazolium, pyridinium, phosphonium and ammonium cations. The use of these novel liquid/liquid biphases for separation of actinides or other fission products associated with nuclear wastes (e.g., pertechnetate salts) has been demonstrated and will be described in this presentation.

  19. Recent developments for an active UF6 gas target for photon-induced fission experiments

    NASA Astrophysics Data System (ADS)

    Freudenberger, M.; Eckardt, C.; Enders, J.; Göök, A.; von Neumann-Cosel, P.; Oberstedt, A.; Oberstedt, S.

    2013-12-01

    Recent developments for an active uranium-hexafluoride-loaded gas target as well as results on the detector gas properties are presented. The gas of choice is a mixture of argon with small amounts of UF6. This contribution presents the experimental setup and focusses on the electron drift velocity with increasing UF6 content. A time-dependent decrease in electron drift velocity is observed in our setup.

  20. 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.

  1. A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

    DOE PAGESBeta

    Knowles, Justin R.; Skutnik, Steven E.; Glasgow, David C.; Kapsimalis, Roger J.

    2016-06-23

    Rapid non-destructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the High Flux Isotope Reactor Neutron Activation Analysis laboratory has developed a generalized non-destructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and capitalizes off of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a holistic characterization of isotopic identification,more » mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% error have been conducted on standards of 235U and 239Pu as low as 12 nanograms in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 229 nanograms of fissile mass with less than 12% error. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. Furthermore, it is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation sources, and account for increasingly complex sample matrices.« less

  2. Influences of Zr, Ce and Ba fission products on the surface properties of UO2: Atomistic simulations

    NASA Astrophysics Data System (ADS)

    Xiao, Hongxing; Long, Chongsheng; Tian, Xiaofeng; Chen, Hongsheng

    2016-07-01

    Molecular dynamics (MD) simulations with a shell-core model have been carried out to investigate the influences of Zr, Ce and Ba fission products on the surface properties of UO2. Simulation results indicate that (i) the presence of these fission products will change the surface energy of three low-index surfaces in UO2; (ii) the individual addition of Ce has no significant effect on the surface energy, while the individual addition of Ba will dramatically decrease the surface energy of UO2 by approximately 18% on (100) surface, 7% on (110) surface and 9% on (111) surface with the Ba contents ranging from 0 to 12.5 mol% at 300 K, which is obviously contrary to the Zr; (iii) the combined additions of Zr, Ce and Ba fission products will continuously increase the surface energy of UO2 (100), (110) and (111) surfaces; (iv) the structures of the three low-index surfaces in pure UO2 as well as U0.8(Zr, Ce, Ba)0.2O2 are dramatically disturbed after the free relaxation; (v) The nearest O atoms move towards the Zr and Ce atoms center by about 0.21 Å and 0.12 Å but move away from the Ba atom center by about 0.27 Å.

  3. Energetics of gaseous and volatile fission products in molten U-10Zr alloy: A density functional theory study

    NASA Astrophysics Data System (ADS)

    Wang, Ning; Tian, Jie; Jiang, Tao; Yang, Yanqiu; Hu, Sheng; Peng, Shuming; Yan, Liuming

    2015-11-01

    Gaseous and volatile fission products have a number of adverse effects on the safety and efficiency of the U-10Zr alloy fuel. The theoretical calculations were applied to investigate the energetics related to the formation, nucleation, and degassing of gaseous and volatile fission products (Kr, Xe and I) in molten U-10Zr alloy. The molecular dynamics (MD) simulations were applied to generate equilibrium configurations. The density functional theory (DFT) calculations were used to build atomistic models including molten U-10Zr alloy as well as its fission products substituted systems. The vacancy formation in liquid U-10Zr alloy were studied using DFT calculations, with average Gibbs free formation energies at 8.266 and 6.333 eV for U- and Zr-vacancies, respectively. And the interaction energies were -1.911 eV, -2.390 eV, and -1.826 eV for the I-I, Xe-Xe, and Kr-Kr interaction in lattice when two of the adjacent uranium atoms were substituted by gaseous atoms. So it could be concluded that the interaction between I, Kr, and Xe in lattice is powerful than the interaction between these two atoms and the other lattice atoms in U-10Zr.

  4. 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.

  5. Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests

    SciTech Connect

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.

  6. Assessment of fission product content of high-level liquid waste supernate on E-Area vault package criteria

    SciTech Connect

    Brown, D.F.

    1994-06-30

    This report assesses the tank farm`s high level waste supernate to determine any potential impacts on waste certification for the E-Area vaults (EAV). The Waste Acceptance Criteria procedure (i.e., WAC 3.10 of the 1S manual) imposes administrative controls on radioactive material in waste packages sent to the EAV, specifically on six fission products. Waste tank supernates contain various fission products, so any waste package containing material contaminated with supernate will contain these radioactive isotopes. This report develops the process knowledge basis for characterizing the supernate composition for these isotopes, so that appropriate controls can be implemented to ensure that the EAV WAC is met. Six fission products are listed in the SRS 1S Manual WAC 3.10: Se-79, which decays to bromine; Sr-90, which decays to niobium; Tc-99, which decays to ruthenium; Sn-126, which decays to tellurium; I-129, which decays to xenon; and Cs-137, which decays to barium.

  7. Independent analysis of selected core-concrete interaction and fission product release experiments with CORCON-MOD2 and VANESA

    SciTech Connect

    Greene, G.A.; Sanborn, Y.

    1986-01-01

    The discrepancies between experimental findings and the Reactor Safety Study predictions, as well as the rapidly developing data base enabling phenomenological modeling of core-concrete interactions and ex-vessel fission product release, have prompted the development of several new computer models of core-concrete interactions and fission product release during severe accidents. Two such models are the CORCON-MOD2 model of core-concrete interactions and the VANESA model of ex-vessel aerosol and fission product release during core-concrete interactions. The final judge on the adequacy of the development of models of core debris-concrete interactions is, of course, comparison of the model predictions with the results of experiments. The research into ex-vessel core debris behavior differs from research into many aspects of reactor accidents in that there are many experimental results for comparison. Comparisons of code predictions with results of tests using realistic temperatures and conditions should provide an indication of the progress that has been made and, with appropriately chosen tests, an indication of work that needs to be done. 9 refs., 6 figs., 5 tabs.

  8. A GENERALIZED METHOD FOR CHARACTERIZATION OF 235U AND 239PU CONTENT USING SHORT-LIVED FISSION PRODUCT GAMMA SPECTROSCOPY

    SciTech Connect

    Knowles, Justin R; Skutnik, Steven E; Glasgow, David C; Kapsimalis, Roger J

    2016-01-01

    Rapid non-destructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the High Flux Isotope Reactor Neutron Activation Analysis laboratory has developed a generalized non-destructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and capitalizes off of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a holistic characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% error have been conducted on standards of 235U and 239Pu as low as 12 nanograms in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 229 nanograms of fissile mass with less than 12% error. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation sources, and account for increasingly complex sample matrices.

  9. Antrim gas play, production expanding in Michigan

    SciTech Connect

    Not Available

    1994-05-30

    Devonian Antrim shale gas, the Michigan basin's dominant hydrocarbon play in terms of number of wells drilled for several years, shows every sign of continuing at a busy pace. About 3,500 Antrim completions now yield 350 MMcfd, more than 60% of Michigan's gas production. The outlook is for Antrim production to climb in the next 2--3 years to 500--600 MMcfd, about 1% of US gas output. These delivery numbers, slow decline rates, and expected producing life of 20--30 years has snagged pipelines attention. The growing production overtaxed local gathering facilities last fall, and the play recently got its first interstate outlet. Completion and production technology advances are improving well performance and trimming costs. Several hundred wells a year are likely to be drilled during the next few years. Production increases are coming from new wells, deepenings, and workovers. Numerous pipeline/gathering projects are planned in the area to handle the growing Antrim volumes. The paper discusses the development of this resource, efforts to extend the play, geology and production, drilling programs, and gas transportation.

  10. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    SciTech Connect

    Myers, B.F.; Morrissey, R.E.

    1980-04-01

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO/sub 2/ and highly enriched (HEU) TRISO UC/sub 2/ particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO/sub 2/ particles and 23.5 and 74% FIMA for UC/sub 2/ particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium.

  11. Fission products silver, palladium, and cadmium identification in neutron-irradiated SiC TRISO particles using a Cs-Corrected HRTEM

    NASA Astrophysics Data System (ADS)

    van Rooyen, I. J.; Olivier, E. J.; Neethling, J. H.

    2016-08-01

    Electron microscopy investigations of selected coated particles from the first advanced gas reactor experiment at Idaho National Laboratory provided important information on fission product distribution and chemical composition in the silicon-carbide (SiC) layer. Silver precipitates were nano-sized, and therefore high-resolution transmission electron microscopy (HRTEM) was used to provide more information at the atomic level. Based on gamma-ray analysis, this particle which was irradiated to an average burnup of 19.38% fissions per initial metal atom, may have released as much as 10% of its available Ag-110 m inventory during irradiation. The HRTEM investigation focused on silver, palladium, and cadmium due to interest in silver transport mechanisms and possible correlation with palladium and silver previously found. Palladium, silver, and cadmium were found to co-exist in some of the SiC grain boundaries and triple junctions. This study confirmed palladium both at inter and intragranular sites. Phosphor was identified in SiC grain boundaries and triple points.

  12. 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.

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

    SciTech Connect

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

    2014-06-15

    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.

  14. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    SciTech Connect

    Mueller, Don; Rearden, Bradley T; Reed, Davis Allan

    2010-01-01

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

  15. NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION

    SciTech Connect

    OH,S.Y.; CHANG,J.; MUGHABGHAB,S.

    2000-05-11

    Neutron cross section evaluations of the fission-product isotopes, {sup 95}Mo, {sup 99}Tc, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, {sup 141}Nd, {sup 147}Sm, {sup 149}Sm, {sup 150}Sm, {sup 151}Sm, {sup 152}Sm, {sup 153}Eu, {sup 155}Gd, and {sup 157}Gd were carried out below the fast neutron energy region within the framework of the BNL-KAERI international collaboration. In the thermal energy region, the energy dependence of the various cross-sections was calculated by applying the multi-level Breit-Wigner formalism. In particular, the strong energy dependence of the coherent scattering lengths of {sup 155}Gd and {sup 157}Gd were determined and were compared with recent calculations of Lynn and Seeger. In the resonance region, the recommended resonance parameters, reported in the BNL compilation, were updated by considering resonance parameter information published in the literature since 1981. The s-wave and, if available, p-wave reduced neutron widths were analyzed in terms of the Porter-Thomas distribution to determine the average level spacings and the neutron strength functions. Average radiative widths were also calculated from measured values of resolved energy resonances. The average resonance parameters determined in this study were compared with those in the BNL and other compilations, as well as the ENDF/B-VI, JEF-2.2, and JENDL-3.2 data libraries. The unresolved capture cross sections of these isotopes, computed with the determined average resonance parameters, were compared with measurements, as well as the ENDF/B-VI evaluations. To achieve agreement with the measurements, in a few cases minor adjustments in the average resonance parameters were made. Because of astrophysical interest, the Maxwellian capture cross sections of these nuclides at a neutron temperature of 30 keV were computed and were compared with other compilations and evaluations.

  16. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

    Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

  17. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    SciTech Connect

    Sforza, P.M.; Cresci, R.J.

    1997-05-31

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical.

  18. Analyzing Losses: Transuranics into Waste and Fission Products into Recycled Fuel

    SciTech Connect

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert E. Cherry; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros; Candido Pereira; Denia Djokic

    2010-11-01

    All mass streams from separations and fuel fabrication are products that must meet criteria. Those headed for disposal must meet waste acceptance criteria (WAC) for the eventual disposal sites corresponding to their waste classification. Those headed for reuse must meet fuel or target impurity limits. A “loss” is any material that ends up where it is undesired. The various types of losses are linked in the sense that as the loss of transuranic (TRU) material into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. We have analyzed four separation options and two fuel fabrication options in a generic fuel cycle. The separation options are aqueous uranium extraction plus (UREX+1), electrochemical, Atomics International reduction oxidation separation (AIROX), and melt refining. UREX+1 and electrochemical are traditional, full separation techniques. AIROX and melt refining are taken as examples of limited separations, also known as minimum fuel treatment. The fuels are oxide and metal. To define a generic fuel cycle, a fuel recycling loop is fed from used light water reactor (LWR) uranium oxide fuel (UOX) at 51 MWth-day/kg-iHM burnup. The recycling loop uses a fast reactor with TRU conversion ratio (CR) of 0.50. Excess recovered uranium is put into storage. Only waste, not used fuel, is disposed – unless the impurities accumulate to a level so that it is impossible to make new fuel for the fast reactor. Impurities accumulate as dictated by separation removal and fission product generation. Our model approximates adjustment to fast reactor fuel stream blending of TRU and U products from incoming LWR UOX and recycling FR fuel to compensate for impurity accumulation by adjusting TRU:U ratios. Our mass flow model ignores postulated fuel impurity limits; we compare the calculated impurity values with those limits to identify elements of concern. AIROX and melt refining cannot be used to separate used LWR UOX-51 because they cannot

  19. Interconnecting compressors control coalbed gas production

    SciTech Connect

    Payton, R.; Niederhofer, J. )

    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.

  20. Coal or natural gas for ecofuel production

    SciTech Connect

    Geertsema, A.

    1998-07-01

    The paper reviews the technology of the Fischer-Tropsch synthesis used in the Sasal plant in South Africa. It discusses environmental aspects and economics of new FT facilities for the production of diesel fuels. Several projects are briefly described which use this technology for natural gas conversion.

  1. 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.

  2. New Methodology for Natural Gas Production Estimates

    EIA Publications

    2010-01-01

    A new methodology is implemented with the monthly natural gas production estimates from the EIA-914 survey this month. The estimates, to be released April 29, 2010, include revisions for all of 2009. The fundamental changes in the new process include the timeliness of the historical data used for estimation and the frequency of sample updates, both of which are improved.

  3. Recoil-α-fission and recoil-α-α-fission events observed in the reaction 48Ca + 243Am

    NASA Astrophysics Data System (ADS)

    Forsberg, U.; Rudolph, D.; Andersson, L.-L.; Di Nitto, A.; Düllmann, Ch. E.; Fahlander, C.; 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.; Schädel, M.; Yakushev, A.; Åberg, S.; Ackermann, D.; Block, M.; Brand, H.; Carlsson, B. G.; Cox, D.; Derkx, X.; Dobaczewski, J.; Eberhardt, K.; Even, J.; Gerl, J.; Jäger, 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, Yue; Thörle-Pospiech, P.; Torres, T.; Traut, T.; Trautmann, N.; Türler, A.; Ward, A.; Ward, D. E.; Wiehl, N.

    2016-09-01

    Products of the fusion-evaporation reaction 48Ca + 243Am were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum für Schwerionenforschung, Darmstadt, Germany. Amongst the detected thirty correlated α-decay chains associated with the production of element Z = 115, two recoil-α-fission and five recoil- α- α-fission events were observed. The latter five chains are similar to four such events reported from experiments performed at the Dubna gas-filled separator, and three such events reported from an experiment at the Berkeley gas-filled separator. The four chains observed at the Dubna gas-filled separator were assigned to start from the 2n-evaporation channel 289115 due to the fact that these recoil- α- α-fission events were observed only at low excitation energies. Contrary to this interpretation, we suggest that some of these recoil- α- α-fission decay chains, as well as some of the recoil- α- α-fission and recoil-α-fission decay chains reported from Berkeley and in this article, start from the 3n-evaporation channel 288115.

  4. Fission fusion hybrids- recent progress

    NASA Astrophysics Data System (ADS)

    Kotschenreuther, M.; Valanju, P.; Mahajan, S.; Covele, B.

    2012-03-01

    Fission-fusion hybrids enjoy unique advantages for addressing long standing societal acceptability issues of nuclear fission power, and can do this at a much lower level of technical development than a competitive fusion power plant- so it could be a nearer term application. For waste incineration, hybrids can burn intransigent transuranic residues (with the long lived biohazard) from light water reactors (LWRs) with far fewer hybrid reactors than a comparable system within the realm of fission alone. For fuel production, hybrids can produce fuel for ˜4 times as many LWRs with NO fuel reprocessing. For both waste incineration or fuel production, the most severe kind of nuclear accident- runaway criticality- can be excluded, unlike either fast reactors or typical accelerator based reactors. The proliferation risks for hybrid fuel production are, we strongly believe, far less than any other fuel production method, including today's gas centrifuges. US Thorium reserves could supply the entire US electricity supply for centuries. The centerpiece of the fuel cycle is a high power density Compact Fusion Neutron Source (major+minor radius ˜ 2.5-3.5 m), which is made feasible by the super-X divertor.

  5. Fission Product Yields for 14 MeV Neutrons on 235U, 238U and 239Pu

    NASA Astrophysics Data System (ADS)

    Mac Innes, M.; Chadwick, M. B.; Kawano, T.

    2011-12-01

    We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on 235U, 238U and 239Pu. The results are from historical measurements made in the 1950s-1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for 235U and 238U, but our FPYs are generally higher for 239Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV 239Pu fission cross section is now known to be 15-20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.

  6. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    NASA Astrophysics Data System (ADS)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    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 Mo99 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 (× 106 cm-1) in a tube, their delta reactivities are the still

  7. Effects of radiation and fission product incorporation in a yttria-stabilized zirconia based inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Zhu, Sha

    This work has investigated the irradiation and incorporation effects of fission products in a yttria-stabilized zirconia (YSZ) based inert matrix fuel (IMF). The concept of inert matrix fuel is based on a new strategy for disposition of plutonium generated from the reprocessing of commercial nuclear fuel and the dismantling of nuclear weapons, i.e. using uranium-free oxides to "burn" plutonium and other actinides (Np, Cm, and Am) in reactors. This approach allows direct disposal, without reprocessing, after once-through burn-up. YSZ and MgAl2O4-YSZ composites are among the potential ceramics for IMF due to their high chemical durability and radiation resistance. The research involved investigating the production, nature, and accumulation of irradiation-induced defects, the behavior of the fission products in the ceramics, the structural stability and amorphization resistance of the YSZ during implantation. Ion implantations were conducted with 200--400 keV Cs+, Sr+, I+, Xe+ and Ti+ up to fluences of 1 x 1017/cm 2 at both room temperature and temperatures of 600--700°C. Thermal annealing was subsequently completed after room temperature ion implantations. In situ and ex situ transmission electron microscopy (TEM), optical absorption spectroscopy, photo-luminescence spectroscopy, and electron paramagnetic resonance (EPR) spectroscopy were employed to characterize the irradiation induced defect evolution and analyze the defect structures. Various irradiation effects were observed and determined in the experiments, such as point defects (F type and V type color centers), defect clusters (dislocation loops), cavities (voids and bubbles), the crystalline-to-amorphous transition, and the phase transformation from fluorite to pyrochlore structure. The ion irradiation-induced amorphization mechanism, the retention ability of the fission products, and structural stability of YSZ are discussed in terms of ion incorporation effects, implanted ion radii, and the solubility

  8. Flibe blanket concept for transmuting transuranic elements and long lived fission products.

    SciTech Connect

    Gohar, Y.

    2000-11-15

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  9. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

    DOE PAGESBeta

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methodsmore » and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the

  10. Microstructural Characterization of Irradiated U-7Mo/Al-5Si Dispersion to High Fission Density

    SciTech Connect

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

    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 work summarizes the recent TEM microstructural characterization of an irradiated U-7Mo/Al-5Si dispersion fuel plate (R3R050) irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory 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 large 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 of the dispersion fuels irradiated to lower fission density from the same fuel plate.

  11. FEMAXI-V benchmarking study on peak temperature and fission gas release prediction of PWR rod fuel

    SciTech Connect

    Suwardi; Dewayatna, W.; Briyatmoko, B.

    2012-06-06

    The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK{center_dot}CEN and Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

  12. 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 Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

  13. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2±x: 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. UO2±x), 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 UO2±x 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 UO2±x 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 Risø fuel rod irradiation experiment was simulated.

  14. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    SciTech Connect

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  15. Use of Information Theory Concepts for Developing Contaminated Site Detection Method: Case for Fission Product and Actinides Accumulation Modeling

    SciTech Connect

    Harbachova, N.V.; Sharavarau, H.A.

    2006-07-01

    Information theory concepts and their fundamental importance for environmental pollution analysis in light of experience of Chernobyl accident in Belarus are discussed. An information and dynamic models of the radionuclide composition formation in the fuel of the Nuclear Power Plant are developed. With the use of code DECA numerical calculation of actinides (58 isotopes are included) and fission products (650 isotopes are included) activities has been carried out and their dependence with the fuel burn-up of the RBMK-type reactor have been investigated. (authors)

  16. Windowless gas targets for neutron production

    NASA Astrophysics Data System (ADS)

    Iverson, Erik. B.; Lanza, Richard C.; Lidsky, L. M.

    1997-02-01

    A windowless deuterium gas target has been constructed for either monoenergetic or white neutron production with a 900 KeV deuteron accelerator. The target is capable of operation at 100 mbar target pressure, and can admit a beam of 5 mm transverse extent. This target is further being modified by the inclusion of an intermittent valve arrangement to reduce the flow rates in the higher pressure stages. This valve should allow operation at up to 1000 mbar with low duty factor beams.

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

    SciTech Connect

    Badamsambuu, J.; Khuukhenkhuu, G.; Norov, N.; Zuzaan, P.; Belov, A. G.; Gangrsky, Yu. P.

    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. Modelling of fission gas release from irradiated UO2 fuel under high-temperature annealing conditions

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Shestak, V. E.

    2012-11-01

    The new model for the vacancy field evolution in grains during annealing of irradiated fuel was developed and implemented in the MFPR code. The model simulates time and spatial variation of the vacancy concentration in the presence of extended vacancy sources (grain boundaries and dislocations) and sinks (growing intragranular bubbles). Being combined with the models for dislocation creep and for bubbles biased migration in the vacancy gradient, the new model self-consistently describes the processes of gas release and microstructure evolution observed in the annealing tests.

  19. Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

    PubMed

    Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

    2015-11-17

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage. PMID:26513644

  20. 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.

  1. Summary Report: Glass-Ceramic Waste Forms for Combined Fission Products

    SciTech Connect

    Crum, Jarrod V.; Riley, Brian J.; Turo, Laura A.; Tang, Ming; Kossoy, Anna

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance

  2. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    SciTech Connect

    Chadwick, M. B.; Herman, Micheal W; Oblozinsky, Pavel; Dunn, Michael E; Danon, Y.; Kahler, A.; Smith, Donald L.; Pritychenko, B; Arbanas, Goran; Arcilla, r; Brewer, R; Brown, D A; Capote, R.; Carlson, A. D.; Cho, Y S; Derrien, Herve; Guber, Klaus H; Hale, G. M.; Hoblit, S; Holloway, Shannon T.; Johnson, T D; Kawano, T.; Kiedrowski, B C; Kim, H; Kunieda, S; Larson, Nancy M; Leal, Luiz C; Lestone, J P; Little, R C; Mccutchan, E A; Macfarlane, R E; MacInnes, M; Matton, C M; Mcknight, R D; Mughabghab, S F; Nobre, G P; Palmiotti, G; Palumbo, A; Pigni, Marco T; Pronyaev, V. G.; Sayer, Royce O; Sonzogni, A A; Summers, N C; Talou, P; Thompson, I J; Trkov, A.; Vogt, R L; Van der Marck, S S; Wallner, A; White, M C; Wiarda, Dorothea; Young, P C

    2011-01-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide

  3. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    NASA Astrophysics Data System (ADS)

    Chadwick, M. B.; Herman, M.; Obložinský, P.; Dunn, M. E.; Danon, Y.; Kahler, A. C.; Smith, D. L.; Pritychenko, B.; Arbanas, G.; Arcilla, R.; Brewer, R.; Brown, D. A.; Capote, R.; Carlson, A. D.; Cho, Y. S.; Derrien, H.; Guber, K.; Hale, G. M.; Hoblit, S.; Holloway, S.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Kim, H.; Kunieda, S.; Larson, N. M.; Leal, L.; Lestone, J. P.; Little, R. C.; McCutchan, E. A.; MacFarlane, R. E.; MacInnes, M.; Mattoon, C. M.; McKnight, R. D.; Mughabghab, S. F.; Nobre, G. P. A.; Palmiotti, G.; Palumbo, A.; Pigni, M. T.; Pronyaev, V. G.; Sayer, R. O.; Sonzogni, A. A.; Summers, N. C.; Talou, P.; Thompson, I. J.; Trkov, A.; Vogt, R. L.; van der Marck, S. C.; Wallner, A.; White, M. C.; Wiarda, D.; Young, P. G.

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range

  4. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    SciTech Connect

    Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  5. Measuring and predicting the transport of actinides and fission product contaminants in unsaturated prairie soil

    NASA Astrophysics Data System (ADS)

    Sims, D. J.

    Soil samples have been taken in 2001 from the area of a 1951 release from an underground storage tank of 6.7 L of an aqueous solution of irradiated uranium (360 GBq). A simulation of the dispersion of the actinides and fission products was conducted in the laboratory using irradiated natural uranium, non-irradiated natural uranium and metal standards dissolved in acidic aqueous solutions and added to soil columns containing uncontaminated prairie soil. The lab soil columns were allowed 12 to 14 months for contaminant transport. Soil samples were analyzed using gamma-ray spectroscopy, neutron activation analysis (NAA) and liquid scintillation counting (LSC) to determine the elemental concentrations of U, Cs and Sr. Diffusion coefficients from the 50 year soil samples and the lab soil samples were determined. The measured diffusion coefficients from the field samples were 3.0 x 10-4 cm2 s-1 (Cs-137), 1.8 x 10-5 cm2 s-1 (U-238) and 2.6 x 10-3 cm2 s-1 (Sr-90) and the values determined from lab simulation were 5 x 10-6 cm 2 s-1 (Cs-137), 3 x 10-5 cm2 s-1 (U-238) and 1.9 x 10-5 cm 2 s-1 (Sr-90). The differences between the sets of diffusion coefficients can be attributed to differences in retardation effects, weather effects and changes in the soil characteristics when transporting, such as porosity. The analytical work showed that Cs-137 content of soil can be determined effectively using gamma-ray spectroscopy; U-238 content can be measured using NAA; and Sr-90 content can be measured using LSC. For non- and low-radioactive species, it was shown that both flame atomic absorption spectrometry (FAAS) and inductively-coupled plasma-mass spectrometry (ICP-MS) gave comparable results for Sr, Cs and Sm, with the average values ranging from 0.5 to 4.5 ppm of each other. The U-238 content results from NAA and from ICP-MS showed general agreement with an average difference of 81.3 ppm on samples having concentrations up to 988.2 ppm. The difference may have been due to matrix

  6. Formation of (Cr, Al)UO4 from doped UO2 and its influence on partition of soluble fission products

    NASA Astrophysics Data System (ADS)

    Cooper, M. W. D.; Gregg, D. J.; Zhang, Y.; Thorogood, G. J.; Lumpkin, G. R.; Grimes, R. W.; Middleburgh, S. C.

    2013-11-01

    CrUO4 and (Cr, Al)UO4 have been fabricated by a sol-gel method, studied using diffraction techniques and modelled using empirical pair potentials. Cr2O3 was predicted to preferentially form CrUO4 over entering solution into hyper-stoichiometric UO2+x by atomic scale simulation. Further, it was predicted that the formation of CrUO4 can proceed by removing excess oxygen from the UO2 lattice. Attempts to synthesise AlUO4 failed, instead forming U3O8 and Al2O3. X-ray diffraction confirmed the structure of CrUO4 and identifies the existence of a (Cr, Al)UO4 phase for the first time (with a maximum Al to Cr mole ratio of 1:3). Simulation was subsequently used to predict the partition energies for the removal of fission products or fuel additives from hyper-stoichiometric UO2+x and their incorporation into the secondary phase. The partition energies are consistent only with smaller cations (e.g. Zr4+, Mo4+ and Fe3+) residing in CrUO4, while all divalent cations are predicted to remain in UO2+x. Additions of Al had little effect on partition behaviour. The reduction of UO2+x due to the formation of CrUO4 has important implications for the solution limits of other fission products as many species are less soluble in UO2 than UO2+x.

  7. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Javed Iqbal, M.; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  8. 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.

  9. Molecular dynamics study of fission gas bubble nucleation in UO2

    NASA Astrophysics Data System (ADS)

    Liu, X.-Y.; Andersson, D. A.

    2015-07-01

    Molecular dynamics (MD) simulations are used to study helium and xenon gas bubble nucleation in UO2. For helium bubbles, the pressure release mechanism is by creating defects on the oxygen sublattice. Helium atoms diffuse away from the bubbles into nearby bulk UO2, thus forming a diffuse interface. For xenon bubbles, over-pressurized bubbles containing xenon can displace uranium atoms, which tend to aggregate around the xenon bubble as a pressure release mechanism. MD simulations of xenon atoms in pre-existing voids suggest that xenon atoms and the replaced uranium atoms occur in a 1:1 ratio, although kinetic factors may reduce that ratio depending on availability of xenon atoms and vacancies around the bubble. Finally, MD simulations suggest that for small bubbles (1-5 xenon atoms), the xenon bubble nucleus at UO2 grain-boundaries has much lower formation energy compared to that of bubbles of similar sizes in the bulk. However, when the xenon bubble grows into larger sizes, this energy difference is reduced.

  10. 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

    SciTech Connect

    Myers, B.F.

    1995-09-01

    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 conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

  11. CO Methanation for Synthetic Natural Gas Production.

    PubMed

    Kambolis, Anastasios; Schildhauer, Tilman J; Kröcher, Oliver

    2015-01-01

    Energy from woody biomass could supplement renewable energy production towards the replacement of fossil fuels. A multi-stage process involving gasification of wood and then catalytic transformation of the producer gas to synthetic natural gas (SNG) represents progress in this direction. SNG can be transported and distributed through the existing pipeline grid, which is advantageous from an economical point of view. Therefore, CO methanation is attracting a great deal of attention and much research effort is focusing on the understanding of the process steps and its further development. This short review summarizes recent efforts at Paul Scherrer Institute on the understanding of the reaction mechanism, the catalyst deactivation, and the development of catalytic materials with benign properties for CO methanation. PMID:26598405

  12. Measurement of fission products β decay properties using a total absorption spectrometer

    NASA Astrophysics Data System (ADS)

    Zakari-Issoufou, A.-A.; Porta, A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Agramunt, J.; Äystö, J.; Bowry, M.; Bui, V. M.; Caballero-Folch, R.; Cano-Ott, D.; Eloma, V.; Estévez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttilä, H.; Regan, P. H.; Rissanen, J.; Rubio, B.; Weber, C.

    2013-12-01

    In a nuclear reactor, the β decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyväskylä with a Total Absorption Spectrometer (TAS) in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

  13. Bio Gas Oil Production from Waste Lard

    PubMed Central

    Hancsók, Jenő; Baladincz, Péter; Kasza, Tamás; Kovács, Sándor; Tóth, Csaba; Varga, Zoltán

    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 = 280–380°C, P = 20–80 bar, LHSV = 0.75–3.0 h−1 and H2/lard ratio: 600 Nm3/m3). In case of the isomerization at the favourable process parameters (T = 360–370°C, 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

  14. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle

    NASA Astrophysics Data System (ADS)

    Fang, Chao; Cao, Jian-Zhu; Sun, Li-Feng

    2011-05-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (UO2) particle is built. The adsorption effect of the fission product on the surface of the UO2 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag—Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor.

  15. Measurements of extinct fission products in nuclear bomb debris: Determination of the yield of the Trinity nuclear test 70 y later.

    PubMed

    Hanson, Susan K; Pollington, Anthony D; Waidmann, Christopher R; Kinman, William S; Wende, Allison M; Miller, Jeffrey L; Berger, Jennifer A; Oldham, Warren J; Selby, Hugh D

    2016-07-19

    This paper describes an approach to measuring extinct fission products that would allow for the characterization of a nuclear test at any time. The isotopic composition of molybdenum in five samples of glassy debris from the 1945 Trinity nuclear test has been measured. Nonnatural molybdenum isotopic compositions were observed, reflecting an input from the decay of the short-lived fission products (95)Zr and (97)Zr. By measuring both the perturbation of the (95)Mo/(96)Mo and (97)Mo/(96)Mo isotopic ratios and the total amount of molybdenum in the Trinity nuclear debris samples, it is possible to calculate the original concentrations of the (95)Zr and (97)Zr isotopes formed in the nuclear detonation. Together with a determination of the amount of plutonium in the debris, these measurements of extinct fission products allow for new estimates of the efficiency and yield of the historic Trinity test. PMID:27382169

  16. 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

  17. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    SciTech Connect

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    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

  18. Fission fragment driven neutron source

    DOEpatents

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  19. A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.

    Energy Science and Technology Software Center (ESTSC)

    1983-03-23

    Version 01/02 The code reads multigroup cross sections from a compatible data file and collapses user-selected reaction cross sections to any few-group structure using one of a variety of user neutron flux spectrum options given below: Option Flux description 1 Built-in function including Maxwellian, fission, fusion and slowing-down regions and requiring user-specified parameters and energy-region boundaries. 2 Set of log-log flux-energy interpolation points read from input cross-section data file. 3 Set of log-log flux-energy interpolationmore » points read from user-supplied card input. 4 - 6 Histogram flux values read from user-supplied card input in arbitrary group structure in units of flux-per unit-energy, flux-per-unit lethargy, or integral group flux. LAFPX-E may be used to collapse any set of multigroup reaction cross sections furnished in the required format. However, the code was developed for, and is furnished with, a library of 154-group fission-product cross sections processed from ENDF/B-IV with a typical light water reactor (LWR) flux spectrum and temperature. Four-group radiative capture cross sections produced for LWR calculations are tabulated in the code documentation and are incorporated in the EPRI-CINDER data library, RSIC Code Package CCC-309.« less

  20. 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.

  1. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

    SciTech Connect

    Scaglione, John M; Mueller, Don; Wagner, John C

    2011-01-01

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission

  2. 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

  3. Deep Atomic Binding (DAB) Hypothesis: A New Approach of Fission Product Chemistry

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    Former studies assumed that, after fission process occurs, the highly ionized new born atoms (20-22 positive charge), ionize the media in which they pass through before becoming stable atoms in a manner similar to 4-MeV ?-particles. Via ordinary chemical reactions with the surroundings, each stable atom has a probability to form chemical compound. Since there are about 35 different elemental atoms created through fission processes, a large number of chemical species were suggested to be formed. But, these suggested chemical species were not found in the environment after actual releases of FP during accidents like TMI (USA, 1979), and Chernobyl (former USSR, 1986), also the models based on these suggested reactions and species could not interpret the behavior of these actual species. It is assumed here that the ionization states of the new born atoms and the long term high temperature were not dealt with in an appropriate way and they were the reasons of former models failure. Our new approach of Deep Atomic Binding (DAB) based on the following: 1-The new born atoms which are highly ionized, 10-12 electrons associated with each nucleus, having a large probability to create bonds between them to form molecules. These bonds are at the L, or M shells, and we call it DAB. 2-The molecules stay in the reactor at high temperatures for long periods, so they undergo many stages of composition and decomposition to form giant molecules. By applying DAB approach, field data from Chernobyl, TMI and nuclear detonations could be interpreted with a wide coincidence resulted. (author)

  4. Gas-well production decline in multiwell reservoirs

    SciTech Connect

    Aminian, K.; Ameri, S. ); Stark, J.J. ); Yost, A.B. II )

    1990-12-01

    This paper introduces a pseudosteady-state constant-pressure solution for gas wells. The solution was used to develop a type-curve-based method to history match and predict multiwell gas reservoir production. Good agreements between the predicted and actual gas well production rates were obtained.

  5. 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

    2015-01-01

    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

  6. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    SciTech Connect

    Liu, Xiang-Yang; Cooper, Michael William Donald; Mcclellan, Kenneth James; Lashley, Jason Charles; Byler, Darrin David; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  7. 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.

  8. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    PubMed

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján; Rypar, Vojtěch; Koleška, 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

  9. 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.

  10. Spontaneous Fission

    DOE R&D Accomplishments Database

    Segre, Emilio

    1950-11-22

    The first attempt to discover spontaneous fission in uranium was made by [Willard] Libby, who, however, failed to detect it on account of the smallness of effect. In 1940, [K. A.] Petrzhak and [G. N.] Flerov, using more sensitive methods, discovered spontaneous fission in uranium and gave some rough estimates of the spontaneous fission decay constant of this substance. Subsequently, extensive experimental work on the subject has been performed by several investigators and will be quoted in the various sections. [N.] Bohr and [A.] Wheeler have given a theory of the effect based on the usual ideas of penetration of potential barriers. On this project spontaneous fission has been studied for the past several years in an effort to obtain a complete picture of the phenomenon. For this purpose the spontaneous fission decay constants {lambda} have been measured for separated isotopes of the heavy elements wherever possible. Moreover, the number {nu} of neutrons emitted per fission has been measured wherever feasible, and other characteristics of the spontaneous fission process have been studied. This report summarizes the spontaneous fission work done at Los Alamos up to January 1, 1945. A chronological record of the work is contained in the Los Alamos monthly reports.

  11. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  12. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    SciTech Connect

    Henager, Charles H.; Jiang, Weilin

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

  13. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    measured in these areas remained low, ˜5%. The intermediate phase is a release area with moderate release. The average fraction of gas measured in the bubbles was around 20%. In the 235U-rich agglomerates, the Xe SIMS total was very low. This part is a release area. Sharp transitions between initial microstructure and the HBS, often inside one grain. Increase in the resulting grain size with increasing distance from the pellet periphery. The grain sizes are in fact consistent with the MOX measurements [2]. Increase in the bubble size with the increasing distance from the pellet periphery, consistent with the MOX measurements. Smaller bubbles tend to be found in the peripheral part of the 235U-rich agglomerates rather than in their central part. Sharp transition, around 0.5R, between the peripheral area where the conventional form of HBS forms in the 235U-rich agglomerates and the central part where much larger bubbles form and where the grain size is also clearly larger. Xe concentration of 0.1 wt% outside the bubbles in the HBS areas is consistent with the [2] MOX measurements at equivalent local burn-ups. The heterogeneous MOX fuels examinations have firmly established that the HBS can extend outside the Pu-rich agglomerates due to the implantation of fission products around these agglomerates. Similarly, it has been shown that the small Pu-rich agglomerates can remain with the initial microstructure even if there is a similar actual local burn-up, a large rate of fission products being implanted outside the agglomerates themselves so that the local fission product concentration remains low.In this 235U heterogeneous UO2, the Xe and Nd concentration levels reached at the HBS formation limit ranged between 0.8 wt% and 1.1 wt% for Xe and between 0.63 wt% and 0.83 wt% for Nd. These ranges are similar to what was reported in [23] for the UO2 rim. These limits are, however, slightly higher than those found for Pu-rich agglomerates in heterogeneous MOX fuels in [2] or in [24

  14. 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

  15. 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

  16. 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

  17. ''Subthreshold'' reactions involving nuclear fission

    SciTech Connect

    Goldhaber, M.; Shrock, R.

    2001-02-01

    We analyze reactions of several types that are naively below threshold but can proceed because of the release of binding energy from nuclear fission and occasionally the formation of Coulombic bound states. These reactions include (i) photofission with pion production and (ii) charged current neutrino-nucleus reactions that lead to fission and/or formation of a Coulomb bound state of a {mu}{sup -} with the nucleus of a fission fragment. We comment on the possible experimental observation of these reactions.

  18. Mobile neutron/gamma waste assay system for characterization of waste containing transuranics, uranium, and fission/activation products

    SciTech Connect

    Davidson, D.R.; Haggard, D.; Lemons, C.

    1994-12-31

    A new integrated neutron/gamma assay system has been built for measuring 55-gallon drums at Pacific Northwest Laboratory. The system is unique because it allows simultaneous measurement of neutrons and gamma-rays. This technique also allows measurement of transuranics (TRU), uranium, and fission/activation products, screening for shielded Special Nuclear Material prior to disposal, and critically determinations prior to transportation. The new system is positioned on a platform with rollers and installed inside a trailer or large van to allow transportation of the system to the waste site instead of movement of the drums to the scanner. The ability to move the system to the waste drums is particularly useful for drum retrieval programs common to all DOE sites and minimizes transportation problems on the site. For longer campaigns, the system can be moved into a facility. The mobile system consists of two separate subsystems: a passive Segmented Gamma Scanner (SGS) and a {open_quotes}clam-shell{close_quotes} passive neutron counter. The SGS with high purity germanium detector and {sup 75}Se transmission source simultaneously scan the height of the drum allowing identification of unshieled {open_quotes}hot spots{close_quotes} in the drum or segments where the matrix is too dense for the transmission source to penetrate. Dense segments can flag shielding material that could be used to hide plutonium or uranium during the gamma analysis. The passive nuetron counter with JSR-12N Neutron Coincidence Analyzer measures the coincident neutrons from the spontaneous fission of even isotopes of plutonium. Because high-density shielding produces minimal absorption of neutrons, compared to gamma rays, the passive neutron portion of the system can detect shielded SNM. Measurements to evaluate the performance of the system are still underway at Pacific Northwest Laboratory.

  19. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    SciTech Connect

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  20. 21 CFR 173.350 - Combustion product gas.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... FOR HUMAN CONSUMPTION (CONTINUED) SECONDARY DIRECT FOOD ADDITIVES PERMITTED IN FOOD FOR HUMAN CONSUMPTION Specific Usage Additives § 173.350 Combustion product gas. The food additive combustion...

  1. Effects of gas bubble production on heat transfer from a volumetrically heated liquid pool

    NASA Astrophysics Data System (ADS)

    Bull, Geoffrey R.

    Aqueous solutions of uranium salts may provide a new supply chain to fill potential shortfalls in the availability of the most common radiopharmaceuticals currently in use worldwide, including Tc99m which is a decay product of Mo99. The fissioning of the uranium in these solutions creates Mo99 but also generates large amounts of hydrogen and oxygen from the radiolysis of the water. When the dissolved gases reach a critical concentration, bubbles will form in the solution. Bubbles in the solution affect both the fission power and the heat transfer out of the solution. As a result, for safety and production calculations, the effects of the bubbles on heat transfer must be understood. A high aspect ratio tank was constructed to simulate a section of an annulus with heat exchangers on the inner and outer steel walls to provide cooling. Temperature measurements via thermocouples inside the tank and along the outside of the steel walls allowed the calculation of overall and local heat transfer coefficients. Different air injection manifolds allowed the exploration of various bubble characteristics and patterns on heat transfer from the pool. The manifold type did not appear to have significant impact on the bubble size distributions in water. However, air injected into solutions of magnesium sulfate resulted in smaller bubble sizes and larger void fractions than those in water at the same injection rates. One dimensional calculations provide heat transfer coefficient values as functions of the superficial gas velocity in the pool.

  2. Influence of the incident particle energy on the fission product mass distribution.

    SciTech Connect

    Gomes, I. C.

    1998-08-26

    For {sup 238}U targets and the five elements considered here, the best yields of neutron-rich isotopes are obtained from neutrons in the 2-20 MeV range. High energy beams of neutrons, protons, and deuterons have comparable integral yields per element to neutrons below 20 MeV, but the distributions are peaked at lower neutron numbers. This is presumably due to a higher neutron multiplicity in the pre-equilibrium stage and/or the compound nucleus/fission stage. For {sup 235}U targets there are high yields predicted especially for thermal neutrons, and also for the fast neutron spectrum. For the high energy neutrons, protons, and deuterons {sup 235}U has no advantage over {sup 238}U. A detailed comparison of the relative advantages of {sup 235}U and {sup 238}U for radioactive beam applications is beyond the scope of this study and will be addressed in the future. The present work is the first step of a more detailed analysis of various possible one- and two-step target geometry calculated with the LAHET code system. It is intended to serve as a guide in choosing geometry and beams for future studies. It is desirable to extend this study to higher beam energies, e.g. 200 to 1000 MeV, but at this time there is very little data against which to benchmark the analysis. Additional data would also permit comparisons of isotope yields beyond the tails of the distributions presented here, to even more neutron rich isotopes.

  3. Fission-fragment excited xenon/rare gas mixtures. II. Small signal gain of the 2. 03 [mu]m xenon transition

    SciTech Connect

    Hebner, G.A.; Hays, G.N. )

    1993-04-15

    The results of small signal gain measurements of the 2.03 [mu]m (5[ital d][3/2][sub 1][minus]6[ital p][3/2][sub 1]) xenon transition in fission-fragment excited Ar/Xe, He/Ar/Xe, Ne/Ar/Xe, and He/Ne/Ar/Xe gas mixtures is presented. Time resolved small signal gain was probed using a cw He/Xe discharge laser as a function of total pressure, xenon concentration, pump power, He/Ne/Ar buffer ratio, and impurity concentration. Small signal gains of up to 6%/cm were observed for pump rates of 15 W/cm[sup 3]. Addition of helium and/or neon to the argon buffer increased the width of the laser gain and reduced the absorption observed under some experimental conditions. Experimentally determined gain scaling laws for several gas mixtures are presented.

  4. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (Inventor)

    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.

  5. 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.

  6. Production of biodiesel using expanded gas solvents

    SciTech Connect

    Ginosar, Daniel M; Fox, Robert V; Petkovic, Lucia M

    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.

  7. Isotopic production cross sections and recoil velocities of spallation-fission fragments in the reaction {sup 238}U(1A GeV)+d

    SciTech Connect

    Pereira, J.; Benlliure, J.; Casarejos, E.; Armbruster, P.; Enqvist, T.; Schmidt, K.-H.; Bernas, M.; Mustapha, B.; Rejmund, F.; Stephan, C.; Taieeb, J.; Tassan-Got, L.; Boudard, A.; Legrain, R.; Leray, S.; Volant, C.; Wlazlo, W.

    2007-01-15

    Fission fragments of 1A GeV{sup 238}U nuclei interacting with a deuterium target have been investigated with the Fragment Separator (FRS) at Gesellschaft fuer Schwerionenforschung (GSI) by measuring their isotopic production cross sections and velocities. Results, along with those obtained recently for spallation-evaporation fragments, provide a comprehensive analysis of the spallation nuclear productions in this reaction. Details about the experimental performance, data reduction and results are presented.

  8. 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.

  9. Windowless gas targets for neutron production

    NASA Astrophysics Data System (ADS)

    Iverson, Erik B.

    A windowless deuterium gas target has been constructed for high yield production of either monoenergetic or white fast neutrons. The operation of this target has been demonstrated on a 900 keV deuteron accelerator. The target is capable of operation at 100 mbar target pressure, and can admit a low duty factor beam of 5 mm transverse extent. The target employs an intermittent valve arrangement to reduce the flow rates in the higher pressure stages of a differentially pumped vacuum system. This valve allows operation at much greater target pressures for low duty factor beams than would otherwise be the case. Neutron yield measurements validated the functionality of the target system. This target will make possible considerable advances in methods of non-destructive testing and evaluation which employ fast neutrons, whether mono-energetic or otherwise. It is further suited to use as a thermal neutron source, with the addition of an appropriate moderator. The development of this target system has not only provided a functioning and valuable piece of equipment for use in further research, but has also investigated the technological limitations and functional requirements of implementing such a system in a practical setting. (Copies available exclusively from MIT Libraries, Rm. 14- 0551, Cambridge, MA 2139-4307. Ph. 617-253-5668; Fax 617- 253-1690.)

  10. 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; Suzuki, Keiji; Yamashita, Shunichi

    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.

  11. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina; Wagner, John C

    2011-01-01

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  12. Spallation reaction study for fission products in nuclear waste: Cross section measurements for 137Cs and 90Sr on proton and deuteron

    NASA Astrophysics Data System (ADS)

    Wang, H.; Otsu, H.; Sakurai, H.; Ahn, D. S.; Aikawa, M.; Doornenbal, P.; Fukuda, N.; Isobe, T.; Kawakami, S.; Koyama, S.; Kubo, T.; Kubono, S.; Lorusso, G.; Maeda, Y.; Makinaga, A.; Momiyama, S.; Nakano, K.; Niikura, M.; Shiga, Y.; Söderström, P.-A.; Suzuki, H.; Takeda, H.; Takeuchi, S.; Taniuchi, R.; Watanabe, Ya.; Watanabe, Yu.; Yamasaki, H.; Yoshida, K.

    2016-03-01

    We have studied spallation reactions for the fission products 137Cs and 90Sr for the purpose of nuclear waste transmutation. The spallation cross sections on the proton and deuteron were obtained in inverse kinematics for the first time using secondary beams of 137Cs and 90Sr at 185 MeV/nucleon at the RIKEN Radioactive Isotope Beam Factory. The target dependence has been investigated systematically, and the cross-section differences between the proton and deuteron are found to be larger for lighter spallation products. The experimental data are compared with the PHITS calculation, which includes cascade and evaporation processes. Our results suggest that both proton- and deuteron-induced spallation reactions are promising mechanisms for the transmutation of radioactive fission products.

  13. 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

  14. Measuring micro-organism gas production

    NASA Technical Reports Server (NTRS)

    Wilkins, J. R.; Pearson, A. O.; Mills, S. M.

    1973-01-01

    Transducer, which senses pressure buildup, is easy to assemble and use, and rate of gas produced can be measured automatically and accurately. Method can be used in research, in clinical laboratories, and for environmental pollution studies because of its ability to detect and quantify rapidly the number of gas-producing microorganisms in water, beverages, and clinical samples.

  15. Mathematical analysis of intermittent gas injection model in oil production

    NASA Astrophysics Data System (ADS)

    Tasmi, Silvya, D. R.; Pudjo, S.; Leksono, M.; Edy, S.

    2016-02-01

    Intermittent gas injection is a method to help oil production process. Gas is injected through choke in surface and then gas into tubing. Gas forms three areas in tubing: gas column area, film area and slug area. Gas column is used to propel slug area until surface. A mathematical model of intermittent gas injection is developed in gas column area, film area and slug area. Model is expanding based on mass and momentum conservation. Using assume film thickness constant in tubing, model has been developed by Tasmi et. al. [14]. Model consists of 10 ordinary differential equations. In this paper, assumption of pressure in gas column is uniform. Model consist of 9 ordinary differential equations. Connection of several variables can be obtained from this model. Therefore, dynamics of all variables that affect to intermittent gas lift process can be seen from four equations. To study the behavior of variables can be analyzed numerically and mathematically. In this paper, simple mathematically analysis approach is used to study behavior of the variables. Variables that affect to intermittent gas injection are pressure in upstream valve and in gas column. Pressure in upstream valve will decrease when gas mass in valve greater than gas mass in choke. Dynamic of the pressure in the gas column will decrease and increase depending on pressure in upstream valve.

  16. Analysis of eastern Devonian gas shales production data

    SciTech Connect

    Gatens, J.M.; Stanley, D.K.; Lancaster, D.E.; Lee, W.J.; Lane, H.S.; Watson, A.T.

    1989-05-01

    Production data from more than 800 Devonian shale wells have been analyzed. Permeability-thickness product and gas in place estimated from production data have been found to correlate with well performance. Empirical performance equations, production type curves, and an analytical dual-porosity model with automatic history-matching scheme were developed for the Devonian shale.

  17. METHOD OF SEPARATING FISSION PRODUCTS FROM FUSED BISMUTH-CONTAINING URANIUM

    DOEpatents

    Wiswall, R.H.

    1958-06-24

    A process is described for removing metal selectively from liquid metal compositions. The method effects separation of flssion product metals selectively from dilute solution in fused bismuth, which contains uraniunn in solution without removal of more than 1% of the uranium. The process comprises contacting the fused bismuth with a fused salt composition consisting of sodium, potassium and lithium chlorides, adding to fused bismuth and molten salt a quantity of bismuth chloride which is stoichiometrically required to convert the flssion product metals to be removed to their chlorides which are more stable in the fused salt than in the molten metal and are, therefore, preferentially taken up in the fused salt phase.

  18. 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

  19. Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

    SciTech Connect

    JAck D. Law

    2007-09-01

    The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

  20. On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Trummer, Martin; Nilsson, Sara; Jonsson, Mats

    2008-08-01

    Radiation induced oxidative dissolution of UO 2 is a key process for the safety assessment of future geological repositories for spent nuclear fuel. This process is expected to govern the rate of radionuclide release to the biosphere. In this work, we have studied the catalytic effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel. The experimental studies were performed using UO 2 pellets containing 0%, 0.1%, 1% and 3% Pd as a model for spent nuclear fuel. H 2O 2 was used as a model for radiolytical oxidants (previous studies have shown that H 2O 2 is the most important oxidant in such systems). The pellets were immersed in aqueous solution containing H 2O 2 and HCO3- and the consumption of H 2O 2 and the dissolution of uranium were analyzed as a function of H 2 pressure (0-40 bar). The noble metal inclusions were found to catalyze oxidation of UO 2 as well as reduction of surface bound oxidized UO 2 by H 2. In both cases the rate of the process increases with increasing Pd content. The reduction process was found to be close to diffusion controlled. This process can fully account for the inhibiting effect of H 2 observed in several studies on spent nuclear fuel dissolution.

  1. Development of a multi-layer diffusion couple to study fission product transport in β-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 β-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.

  2. 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.

  3. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    NASA Astrophysics Data System (ADS)

    Bürger, S.; Riciputi, L. R.; Bostick, D. A.; Turgeon, S.; McBay, E. H.; Lavelle, M.

    2009-09-01

    A ThermoFisher "Triton" multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotope ratio analysis 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 (104 atoms to 105 atoms) for 239-242+244Pu, 233+236U, 241-243Am, 89,90Sr, and 134,135,137Cs, and <=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 × 106 or better using a SEM are reported here. Precisions of RSD [approximate]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.

  4. Depressurization and electrical heating of hydrate sediment for gas production

    NASA Astrophysics Data System (ADS)

    Minagawa, H.

    2015-12-01

    As a part of a Japanese National hydrate research program (MH21, funded by METI), we performed a study on electrical heating of the hydrate core combined with depressurization for gas production. In-situ dissociation of natural gas hydrate is necessary for commercial recovery of natural gas from natural gas hydrate sediment. Thermal stimulation is an effective dissociation method, along with depressurization.To simulate methane gas production from methane hydrate layer, we investigated electrical heating of methane hydrate sediment. A decrease in core temperature due to the endothermic reaction of methane hydrate dissociation was suppressed and the core temperature increased between 1oC and 4oC above the control temperature with electric heating. A current density of 10A/m2 with depressurization would effectively dissociate hydrate. Therefore, depressurization and additional electrode heating of hydrate sediment saturated with electrolyte solution was confirmed to enable higher gas production from sediment with less electric power.

  5. Integrated production of fuel gas and oxygenated organic compounds from synthesis gas

    DOEpatents

    Moore, Robert B.; Hegarty, William P.; Studer, David W.; Tirados, Edward J.

    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.

  6. GGA+U study of uranium mononitride: A comparison of the U-ramping and occupation matrix schemes and incorporation energies of fission products

    NASA Astrophysics Data System (ADS)

    Claisse, Antoine; Klipfel, Marco; Lindbom, Niclas; Freyss, Michel; Olsson, Pär

    2016-09-01

    Uranium mononitride is studied in the DFT + U framework. Its ground state is investigated and a study of the incorporation of diverse fission products in the crystal is conducted. The U-ramping and occupation matrix control (OMC) schemes are used to eliminate metastable states. Beyond a certain amount of introduced correlation, the OMC scheme starts to find a lower total energy. The OMC scheme is chosen for the second part of this study. Furthermore, the influence of the magnetic ordering is studied using the U-ramping method, showing that antiferromagnetic order is the most stable one when the U parameter is larger than 1.75 eV. The effect on the density of states is investigated and elastic constants are provided for comparison with other methods and experiments. The incorporation energies of fission products in different defect configurations are calculated and these energies are corrected to take into account the limited size of the supercell.

  7. 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.

  8. Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

    NASA Astrophysics Data System (ADS)

    Sakama, M.; Nagano, Y.; Kitade, T.; Shikino, O.; Nakayama, S.

    2014-06-01

    Radioactive fission product 131I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, 134Cs and 137Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m-3 in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of 134Cs and 137Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m-3) variation of stable cesium (133Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

  9. Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

    SciTech Connect

    Sakama, M.; Nagano, Y.; Kitade, T.; Shikino, O.; Nakayama, S.

    2014-06-15

    Radioactive fission product {sup 131}I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, {sup 134}Cs and {sup 137}Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m{sup -3} in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of {sup 134}Cs and {sup 137}Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m{sup -3}) variation of stable cesium ({sup 133}Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

  10. Benchmarking nuclear fission theory

    SciTech Connect

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; Talou, P.

    2015-05-14

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, 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.

  11. Process for production desulfurized of synthesis gas

    DOEpatents

    Wolfenbarger, James K.; Najjar, Mitri S.

    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.

  12. 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.

  13. Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

    SciTech Connect

    Rimshaw, S.J.; Case, F.N.

    1980-01-01

    Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO/sub 3/ ..-->.. N/sub 2/O + 4 CO/sub 2/ + 5 H/sub 2/O. These gases can be scrubbed free of /sup 106/Ru, /sup 131/I, and /sup 99/Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO/sub 4/ volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO/sub 4/ in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO/sub 2/ aerosol having a particle size of 0.5 ..mu... This RuO/sub 2/ aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250/sup 0/C and a moderate volatility of 9.5% at 600/sup 0/C. Volatilities of /sup 99/Tc were generally <1% on calcining acidic or basic wastes at temperatures of 250 to 600/sup 0/C. Data are presented to indicate that /sup 99/Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg /sup 99/Tc being associated with each curie of /sup 137/Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste.

  14. Gas hearth products market fact base. Topical report, January 1996

    SciTech Connect

    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.

  15. Environmental policy and regulatory constraints to natural gas production.

    SciTech Connect

    Elcock, D.

    2004-12-17

    For the foreseeable future, most of the demand for natural gas in the United States will be met with domestic resources. Impediments, or constraints, to developing, producing, and delivering these resources can lead to price increases or supply disruptions. Previous analyses have identified lack of access to natural gas resources on federal lands as such an impediment. However, various other environmental constraints, including laws, regulations, and implementation procedures, can limit natural gas development and production on both federal and private lands. This report identifies and describes more than 30 environmental policy and regulatory impediments to domestic natural gas production. For each constraint, the source and type of impact are presented, and when the data exist, the amount of gas affected is also presented. This information can help decision makers develop and support policies that eliminate or reduce the impacts of such constraints, help set priorities for regulatory reviews, and target research and development efforts to help the nation meet its natural gas demands.

  16. Fission Spectrum

    DOE R&D Accomplishments Database

    Bloch, F.; Staub, H.

    1943-08-18

    Measurements of the spectrum of the fission neutrons of 25 are described, in which the energy of the neutrons is determined from the ionization produced by individual hydrogen recoils. The slow neutrons producing fission are obtained by slowing down the fast neutrons from the Be-D reaction of the Stanford cyclotron. In order to distinguish between fission neutrons and the remaining fast cyclotron neutrons both the cyclotron current and the pusle amplifier are modulated. A hollow neutron container, in which slow neutrons have a lifetime of about 2 milliseconds, avoids the use of large distances. This method results in much higher intensities than the usual modulation arrangement. The results show a continuous distribution of neutrons with a rather wide maximum at about 0.8 MV falling off to half of its maximum value at 2.0 MV. The total number of netrons is determined by comparison with the number of fission fragments. The result seems to indicate that only about 30% of the neutrons have energies below .8 MV. Various tests are described which were performed in order to rule out modification of the spectrum by inelastic scattering. Decl. May 4, 1951

  17. Strategies for gas production from oceanic Class 3 hydrateaccumulations

    SciTech Connect

    Moridis, George J.; Reagan, Matthew T.

    2007-05-01

    Gas hydrates are solid crystalline compounds in which gasmolecules are lodged within the lattices of ice crystals. Vast amounts ofCH4 are trapped in gas hydrates, and a significant effort has recentlybegun to evaluate hydrate deposits as a potential energy source. Class 3hydrate deposits are characterized by an isolated Hydrate-Bearing Layer(HBL) that is not in contact with any hydrate-free zone of mobile fluids.The base of the HBL in Class 3 deposits may occur within or at the edgeof the zone of thermodynamic hydrate stability.In this numerical study oflong-term gas production from typical representatives of unfracturedClass 3 deposits, we determine that simple thermal stimulation appears tobe a slow and inefficient production method. Electrical heating and warmwater injection result in very low production rates (4 and 12 MSCFD,respectively) that are orders of magnitude lower than generallyacceptable standards of commercial viability of gas production fromoceanic reservoirs. However, production from depressurization-baseddissociation based on a constant well pressure appears to be a promisingapproach even in deposits characterized by high hydrate saturations. Thisapproach allows the production of very large volumes ofhydrate-originating gas at high rates (>15 MMSCFD, with a long-termaverage of about 8.1 MMSCFD for the reference case) for long times usingconventional technology. Gas production from hydrates is accompanied by asignificant production of water. However, unlike conventional gasreservoirs, the water production rate declines with time. The lowsalinity of the produced water may require care in its disposal. Becauseof the overwhelming advantage of depressurization-based methods, thesensitivity analysis was not extendedto thermal stimulation methods. Thesimulation results indicate that depressurization-induced gas productionfrom oceanic Class 3 deposits increases (and the corresponding waterto-gas ratio decreases) with increasing hydrate temperature

  18. 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.

  19. Outer Continental Shelf Oil and Gas Leasing/Production Program

    SciTech Connect

    Not Available

    1988-01-01

    This annual report on the Outer Continental Shelf (OCS) Oil and Gas Leasing and Production program summarizes receipts and expenditures, and includes information on OCS safety violations as reported by the US Coast Guard. 3 figs., 12 tabs.

  20. Catalyst life and product color prediction for gas oil HDS

    SciTech Connect

    Ushio, M.; Hatayama, M.; Waku, T.

    1995-12-31

    Gas oil hydrodesulfurization was investigated. The sulfur content was reduced by increasing the reaction temperature. However, the severe temperatures made the product oil colored. The kinetic parameters of decoloring reaction at lower tempeatures were calculated.

  1. Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

    NASA Astrophysics Data System (ADS)

    Rajput, M. U.; Ali, N.; Hussain, S.; Mujahid, S. A.; MacMahon, D.

    2012-04-01

    The radionuclide 125Sb is a long-lived fission product, which decays to 125Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

  2. Future challenges for nuclear data research in fission (u)

    SciTech Connect

    Chadwick, Mark B

    2010-01-01

    I describe some high priority research areas in nuclear fission, where applications in nuclear reactor technologies and in modeling criticality in general are demanding higher accuracies in our databases. We focus on fission cross sections, fission neutron spectra, and fission product data.

  3. Mitigating Accidents In Oil And Gas Production Facilities

    NASA Astrophysics Data System (ADS)

    Johnsen, Stig

    Integrated operations are increasingly used in oil and gas production facilities to improve yields, reduce costs and maximize profits. They leverage information and communications technology (ICT) to facilitate collaboration between experts at widely dispersed locations. This paper discusses the safety and security consequences of implementing integrated operations for oil and gas production. It examines the increased accident risk arising from the tight coupling of complex ICT and SCADA systems, and proposes technological, organizational and human factors based strategies for mitigating the risk.

  4. 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.

  5. 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.

  6. Carbonyl Emissions From Oil and Gas Production Facilities

    NASA Astrophysics Data System (ADS)

    Lyman, S. N.; O'Neil, T.; Tran, T.

    2015-12-01

    A number of recent studies have targeted emissions of methane and other hydrocarbons from oil and gas exploration and production activity. These measurements are greatly increasing understanding of the atmospheric impacts of oil and gas development. Very few measurements exist, however, of emissions of formaldehyde and other carbonyls from oil and gas equipment. Carbonyls are toxic and serve as important ozone precursors, especially during winter ozone episodes in places like Utah's Uintah Basin. Current air quality models are only able to reproduce observed high wintertime ozone if they incorporate emissions inventories with very high carbonyl emissions. We measured carbonyl emissions from oil and gas equipment and facilities—including glycol dehydrators, liquid storage tanks, raw gas leaks, raw gas-burning engines, and produced water surface impoundments—in Rocky Mountain oil and gas fields. Carbonyl emissions from raw gas were below detection, but emissions of formaldehyde, acetaldehyde, and other carbonyls were detected from liquid storage tanks, glycol dehydrators, and other oil and gas equipment. In some cases, carbonyls may be formed from the degradation of methanol and other chemicals used in oil and gas production, but the collected data provide evidence for other non-combustion formation pathways. Raw gas-burning engines also emitted carbonyls. Emissions from all measured sources were a small fraction of total volatile organic compound emissions. We incorporated our measurements into an emissions inventory, used that inventory in an air quality model (WRF-SMOKE-CAMx), and were unable to reproduce observed high wintertime ozone. This could be because (1) emission sources we have not yet measured, including compressors, gas processing plants, and others, are large; (2) non-carbonyl emissions, especially those that quickly degrade into carbonyls during photochemical processing, are underestimated in the inventory; or (3) the air quality model is unable

  7. Devices for the Production of Reference Gas Mixtures.

    PubMed

    Fijało, Cyprian; Dymerski, Tomasz; Gębicki, Jacek; Namieśnik, Jacek

    2016-09-01

    For many years there has been growing demand for gaseous reference materials, which is connected with development in many fields of science and technology. As a result, new methodological and instrumental solutions appear that can be used for this purpose. Appropriate quality assurance/quality control (QA/QC) must be used to make sure that measurement data are a reliable source of information. Reference materials are a significant element of such systems. In the case of gas samples, such materials are generally called reference gas mixtures. This article presents the application and classification of reference gas mixtures, which are a specific type of reference materials, and the methods for obtaining them are described. Construction solutions of devices for the production of reference gas mixtures are detailed, and a description of a prototype device for dynamic production of reference gas mixtures containing aroma compounds is presented. PMID:27437588

  8. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    SciTech Connect

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  9. Coal or natural gas for ecofuel production

    SciTech Connect

    Geertsema, A.

    1998-04-01

    Given the extensive available resources of coal and, to a lesser extent, natural gas, the challenge to access these resources in a way that balances growth and conservation in a responsible way, is a tough technological task. On the one hand there is the inadverterable and undesirable liberation of CO{sub 2} when carbon is used and on the other hand it is reasonable to assume that hydrocarbon liquids will, for the foreseeable future, remain the backbone of the supply of energy to automotive vehicles. It is therefore necessary that options for improved environmental performance of such fuels are developed and considered for application where the economics would permit it.

  10. Fission meter

    DOEpatents

    Rowland, Mark S.; Snyderman, Neal J.

    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.

  11. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    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.

  12. 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.

  13. US production of natural gas from tight reservoirs

    SciTech Connect

    Not Available

    1993-10-18

    For the purposes of this report, tight gas reservoirs are defined as those that meet the Federal Energy Regulatory Commission`s (FERC) definition of tight. They are generally characterized by an average reservoir rock permeability to gas of 0.1 millidarcy or less and, absent artificial stimulation of production, by production rates that do not exceed 5 barrels of oil per day and certain specified daily volumes of gas which increase with the depth of the reservoir. All of the statistics presented in this report pertain to wells that have been classified, from 1978 through 1991, as tight according to the FERC; i.e., they are ``legally tight`` reservoirs. Additional production from ``geologically tight`` reservoirs that have not been classified tight according to the FERC rules has been excluded. This category includes all producing wells drilled into legally designated tight gas reservoirs prior to 1978 and all producing wells drilled into physically tight gas reservoirs that have not been designated legally tight. Therefore, all gas production referenced herein is eligible for the Section 29 tax credit. Although the qualification period for the credit expired at the end of 1992, wells that were spudded (began to be drilled) between 1978 and May 1988, and from November 5, 1990, through year end 1992, are eligible for the tax credit for a subsequent period of 10 years. This report updates the EIA`s tight gas production information through 1991 and considers further the history and effect on tight gas production of the Federal Government`s regulatory and tax policy actions. It also provides some high points of the geologic background needed to understand the nature and location of low-permeability reservoirs.

  14. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    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.

  15. 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.

  16. Fission product separations testing

    SciTech Connect

    Bostick, D.A.; DePaoli, S.M.

    1997-10-01

    The initial goal of this task is to adequately understand the treatment needs of the end user in treating contaminated wastewater. These needs are then incorporated into the evaluation of new treatment technologies for wastewater treatment. Pertinent information is than supplied to the end user so that they can select a preferred process to meet their waste treatment needs. New sorbent materials, ion-exchange materials, or other processes of interest to DOE`s Office of Environmental Restoration (EM-40) will be evaluated initially for the removal of {sup 90}Sr and {sup 137}Cs from groundwater and process wastewater. Laboratory studies will strive to obtain a quantitative understanding of the behavior of these new materials and to evaluate their sorption efficiency in reference to a standard benchmark treatment technique. Testing of the new materials will begin by conducting scoping tests where new treatment materials are compared with standard, commercially available materials in batch shaker tests. Experimental data for the most promising sorbents will then be fit to an equilibrium model so that nuclide sorption can be predicted for variable wastewater composition. Additional testing with actual wastewater will be conducted with two or three of the most effective treatment methods. Once batch testing of a treatment method is completed, dynamic column tests will be performed to validate the equilibrium sorption model and to obtain the defining column operating parameters for scaling up the technology.

  17. Water Resources and Natural Gas Production from the Marcellus Shale

    USGS Publications Warehouse

    Soeder, Daniel J.; Kappel, William M.

    2009-01-01

    The Marcellus Shale is a sedimentary rock formation deposited over 350 million years ago in a shallow inland sea located in the eastern United States where the present-day Appalachian Mountains now stand (de Witt and others, 1993). This shale contains significant quantities of natural gas. New developments in drilling technology, along with higher wellhead prices, have made the Marcellus Shale an important natural gas resource. The Marcellus Shale extends from southern New York across Pennsylvania, and into western Maryland, West Virginia, and eastern Ohio (fig. 1). The production of commercial quantities of gas from this shale requires large volumes of water to drill and hydraulically fracture the rock. This water must be recovered from the well and disposed of before the gas can flow. Concerns about the availability of water supplies needed for gas production, and questions about wastewater disposal have been raised by water-resource agencies and citizens throughout the Marcellus Shale gas development region. This Fact Sheet explains the basics of Marcellus Shale gas production, with the intent of helping the reader better understand the framework of the water-resource questions and concerns.

  18. 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.

  19. On the fission chamber pulse charge acquisition and interpretation at MINERVE

    NASA Astrophysics Data System (ADS)

    Loiseau, P.; Geslot, B.; André, J.

    2013-04-01

    Fission Chambers (FCs) are widely used as neutron detectors for online flux measurement. The FC current pulse charge is a key observable quantity which depends on specifications such as the filling gas pressure and the FC geometry. In order to study pulse charges, experimental data have been acquired at the Cadarache zero power reactor MINERVE. Two chambers with contrasting specifications have been used. The experimental pulse charge spectrum is interpreted by the mean of a modeling of fission products (FPs) energy deposition within the filling gas. The pulse charge spectrum peaks are found to correspond to FP emitted perpendicularly to the electrodes.

  20. Gas Production from Hydrate-Bearing Sediments - Emergent Phenomena -

    SciTech Connect

    Jung, J.W.; Jang, J.W.; Tsouris, Costas; Phelps, Tommy Joe; Rawn, Claudia J; Santamarina, Carlos

    2012-01-01

    Even a small fraction of fine particles can have a significant effect on gas production from hydrate-bearing sediments and sediment stability. Experiments were conducted to investigate the role of fine particles on gas production using a soil chamber that allows for the application of an effective stress to the sediment. This chamber was instrumented to monitor shear-wave velocity, temperature, pressure, and volume change during CO{sub 2} hydrate formation and gas production. The instrumented chamber was placed inside the Oak Ridge National Laboratory Seafloor Process Simulator (SPS), which was used to control the fluid pressure and temperature. Experiments were conducted with different sediment types and pressure-temperature histories. Fines migrated within the sediment in the direction of fluid flow. A vuggy structure formed in the sand; these small cavities or vuggs were precursors to the development of gas-driven fractures during depressurization under a constant effective stress boundary condition. We define the critical fines fraction as the clay-to-sand mass ratio when clays fill the pore space in the sand. Fines migration, clogging, vugs, and gas-driven fracture formation developed even when the fines content was significantly lower than the critical fines fraction. These results show the importance of fines in gas production from hydrate-bearing sediments, even when the fines content is relatively low.