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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

    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.

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  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 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.

  19. 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.

  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. 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)

  3. 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.

  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. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  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. (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.

  3. 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.

  4. 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.

  5. 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.

  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. 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.

  8. 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.

  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. 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.

  17. 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.

  18. 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.

  19. 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

  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. 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.

  3. 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.

  4. 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.

  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. 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.

  12. 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.

  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. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

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

    SciTech Connect

    Hoffman, D.C.

    1981-01-01

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

  20. 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.

  1. 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.

  2. 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

  3. 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)

  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. 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

  7. 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.

  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. 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.

  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

    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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  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. 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.

  19. 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.

  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. 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.

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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.

  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. 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.

  12. 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.

  13. 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).

  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. 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)

  17. 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.

  18. 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.

  19. 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.

  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. 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.

  2. 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}

  3. 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

  4. 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.

  5. 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.

  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. 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.

  10. 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.

  11. 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.

  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. 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.

  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. 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.

  20. 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)

  1. 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.

  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. 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.

  10. 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.

  11. 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.

  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. 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

  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. 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)

  17. 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.

  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. 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.

  1. 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)

  2. 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

  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

    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.

  5. 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.

  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. 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)

  10. 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/.

  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. 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.

  15. 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.

  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. 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

  18. 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

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  13. 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.

  14. 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.

  15. 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.

  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. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  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. 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.

  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. 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.

  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. 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.

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

    SciTech Connect

    Bibler, N.E.; Coleman, C.J. ); 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.

  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.

    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.

  18. 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.

  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. 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.

  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. 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.

  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 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.

  7. 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.

  8. 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).

  9. 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.

  10. 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)