Science.gov

Sample records for gas fission products

  1. New Results on Helium and Tritium Gas Production From Ternary Fission

    SciTech Connect

    Serot, O.; Heyse, J.

    2005-05-24

    Ternary fission constitutes an important source of helium and tritium gas production in nuclear reactors and in used fuel elements. Data related to this production are therefore requested by nuclear industry. In the present paper, we report results from measurements of the 4He and 3H emission probabilities (denoted LRA/B and t/B, respectively). These measurements concern both thermal neutron-induced fission reactions as well as spontaneous fission decays. For spontaneous fission, data are reported for nuclides ranging from 238Pu up to 252Cf. For thermal neutron-induced fission, results cover target nuclei between 229Th and 251Cf. Based on these and other results, semi-empirical relations are proposed. These correlations are only valid if spontaneous fission data and neutron-induced fission data are considered separately, which shows the impact of the fissioning nucleus-excitation energy on the ternary particle-emission process. In this way, t/B and LRA/B values could be evaluated for fissioning systems not investigated so far. These results could be used for the ternary fission-yield evaluation of the JEFF3.1 library.

  2. Fission gas detection system

    DOEpatents

    Colburn, Richard P. (Pasco, WA)

    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.

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

    SciTech Connect

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

    2010-10-01

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

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

    SciTech Connect

    Dawn M. Scates; John 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.

  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

    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.

  6. 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.

  7. 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.

  8. FASTGRASS. Fission Product Release Ubase Fuels

    SciTech Connect

    Rest, J.; Zawadzki, S.A.

    1992-09-01

    FASTGRASS is a mechanistic code that predicts atomic and bubble behavior of fission gas in UO2 fuel under steady-state and transient conditions. FASTGRASS also calculates the behavior of the volatile fission products (VFP) I, Cs, and Te as well as the alkaline earth fission products (AEFP) Ba and Sr. The chemistry models include the reaction products Cs2MoO4, Cs2UO4, BaO, SrO and BaUO4. Both the condensed and vapor phases of BaO and SrO are considered. A basic premise of the FASTGRASS analysis is that the noble gases provide the major pathways for fission product release as well as the major reaction sites required for vapor phase formation of various reaction products. Models are included for fission-product generation, atomic migration, bubble nucleation and re-solution, bubble migration and coalescence, channel formation on grain faces, the interlinking of porosity along grain edges, and microcracking on both the amount of fission products released and on their distribution within the fuel. Mechanistic models are also included for grain growth/grain boundary sweeping, and for the behavior of fission products under liquefaction/dissolution and fuel melting conditions. FASTGRASS uses a realistic equation of state for xenon, experimentally derived steady-state bubble mobilities, and phenomenological modeling of bubble mobilities during transient nonequilibrium conditions to calculate the swelling due to retained fission-gas bubbles in the lattice, on grain faces, and along the grain edges. FASTGRASS also calculates fission-product release as a function of time for steady-state and transient thermal conditions. Two variants of the software are included: FASTGRASS, for multinode calculations, and PARAGRASS, for single-node calculations. PARAGRASS can be used as a module in a larger program.

  9. FASTGRASS. Fission Product Release Ubase Fuels

    SciTech Connect

    Rest, J.; Zawadzki, S.A.

    1992-09-01

    FASTGRASS is a mechanistic code that predicts atomic and bubble behavior of fission gas in UO2 fuel under steady-state and transient conditions. FASTGRASS also calculates the behavior of the volatile fission products (VFP) I, Cs, and Te as well as the alkaline earth fission products (AEFP) Ba and Sr. The chemistry models include the reaction products Cs2MoO4, Cs2UO4, BaO, SrO and BaUO4. Both the condensed and vapor phases of BaO and SrO are considered. A basic premise of the FASTGRASS analysis is that the noble gases provide the major pathways for fission product release as well as the major reaction sites required for vapor phase formation of various reaction products. Models are included for fission-product generation, atomic migration, bubble nucleation and re-solution, bubble migration and coalescence, channel formation on grain faces, the interlinking of porosity along grain edges, and microcracking on both the amount of fission products released and on their distribution within the fuel. Mechanistic models are also included for grain growth/grain boundary sweeping, and for the behavior of fission products under liquefaction/dissolution and fuel melting conditions. FASTGRASS uses a realistic equation of state for xenon, experimentally derived steady-state bubble mobilities, and phenomenological modeling of bubble mobilities during transient nonequilibrium conditions to calculate the swelling due to retained fission- gas bubbles in the lattice, on grain faces, and along the grain edges. FASTGRASS also calculates fission-product release as a function of time for steady-state and transient thermal conditions. Two variants of the software are included: FASTGRASS, for multinode calculations, and PARAGRASS, for single-node calculations. PARAGRASS can be used as a module in a larger program.

  10. Reaction Kinetics of a Fission-Product Mixture in a Steam-Hydrogen Carrier Gas in the Phebus Primary Circuit

    SciTech Connect

    Cantrel, Laurent; Krausmann, Elisabeth

    2003-10-15

    Radioiodine entering the containment from the postaccident primary circuit in vapor or gaseous form, as observed in the Phebus FPT0 and FPT1 tests, has a direct impact on the source term evaluation. State-of-the-art fission-product transport codes based on the assumption of thermochemical equilibrium failed to predict this phenomenon. In this work the standard approach of assuming the instantaneous establishment of thermochemical equilibrium is questioned and it will be argued that kinetic limitations may have existed under the severe-accident boundary conditions of the FPT0 and FPT1 tests. To this end a simple monodimensional transport model was developed in an attempt at introducing kinetic aspects within the primary circuit. A number of homogeneous gas-phase reactions between selected fission products and structural materials, complemented by condensation reactions, underlies the kinetic model. In the absence of experimental data, the kinetic constants were estimated using the transition-state theory or semi-empirical methods. The kinetic model was then applied to the analysis of Phebus FPT0 and FPT1 yielding a satisfactory agreement between experimental data and model predictions.

  11. Fission product solvent extraction

    SciTech Connect

    Moyer, B.A.; Bonnesen, P.V.; Sachleben, R.A.

    1998-02-01

    Two main objectives concerning removal of fission products from high-level tank wastes will be accomplished in this project. The first objective entails the development of an acid-side Cs solvent-extraction (SX) process applicable to remediation of the sodium-bearing waste (SBW) and dissolved calcine waste (DCW) at INEEL. The second objective is to develop alkaline-side SX processes for the combined removal of Tc, Cs, and possibly Sr and for individual separation of Tc (alone or together with Sr) and Cs. These alkaline-side processes apply to tank wastes stored at Hanford, Savannah River, and Oak Ridge. This work exploits the useful properties of crown ethers and calixarenes and has shown that such compounds may be economically adapted to practical processing conditions. Potential benefits for both acid- and alkaline-side processing include order-of-magnitude concentration factors, high rejection of bulk sodium and potassium salts, and stripping with dilute (typically 10 mM) nitric acid. These benefits minimize the subsequent burden on the very expensive vitrification and storage of the high-activity waste. In the case of the SRTALK process for Tc extraction as pertechnetate anion from alkaline waste, such benefits have now been proven at the scale of a 12-stage flowsheet tested in 2-cm centrifugal contactors with a Hanford supernatant waste simulant. SRTALK employs a crown ether in a TBP-modified aliphatic kerosene diluent, is economically competitive with other applicable separation processes being considered, and has been successfully tested in batch extraction of actual Hanford double-shell slurry feed (DSSF).

  12. [Fission product yields of 60 fissioning reactions]. Final report

    SciTech Connect

    Rider, B.F.

    1995-05-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ``Evaluation and Compilation of Fission Product Yields 1993,`` LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set.

  13. 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.

  14. Measurement of Fission Product Yields from Fast-Neutron Fission

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  15. 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.

  16. 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 ?m 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 graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted (“Adsorption of Iodine on Graphite in High Temperature Gas-Cooled Reactor Systems: A Review,” Kyle L. Walton, Tushar K. Ghosh, Dabir S. Viswanath, Sudarshan K. Loyalka, Robert V. Tompson). Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.

  17. 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 products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).

  18. Systematics of Fission-Product Yields

    E-print Network

    Wahl, A C

    2002-01-01

    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.

  19. 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)

  20. Electron spectra from decay of fission products

    SciTech Connect

    Dickens, J K

    1982-09-01

    Electron spectra following decay of individual fission products (72 less than or equal to A less than or equal to 162) are obtained from the nuclear data given in the compilation using a listed and documented computer subroutine. Data are given for more than 500 radionuclides created during or after fission. The data include transition energies, absolute intensities, and shape parameters when known. An average beta-ray energy is given for fission products lacking experimental information on transition energies and intensities. For fission products having partial or incomplete decay information, the available data are utilized to provide best estimates of otherwise unknown decay schemes. This compilation is completely referenced and includes data available in the reviewed literature up to January 1982.

  1. 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.

  2. PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA

    E-print Network

    PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA SPECIAL SCIENTIFIC REPORT- FISHERIES No. 167 and Wildlife Service, John L. Farley, Director PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA by Walter A was slow. PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA In relation to "fallout" from nuclear -bomb

  3. 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.

  4. SPIDER Progress Towards High Resolution Correlated Fission Product Data

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

  5. Fission gas release restrictor for breached fuel rod

    DOEpatents

    Kadambi, N. Prasad (Gaithersburg, MD); Tilbrook, Roger W. (Monroeville, PA); Spencer, Daniel R. (Unity Twp., PA); Schwallie, Ambrose L. (Greensburg, PA)

    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.

  6. Nuclear Fission and Fission{minus}Product Spectroscopy: Second International Workshop. Proceedings

    SciTech Connect

    Fioni, G.; Faust, H.; Oberstedt, S.; Hambsch, F.

    1998-10-01

    These proceedings represent papers presented at the Second International Workshop on Nuclear Fission and Fission{minus}Product Spectroscopy held in Seyssins, France in April, 1998. The objective was to bring together the specialists in the field to overview the situation and to assess our present understanding of the fission process. The topics presented at the conference included nuclear waste management, incineration, neutron driven transmutation, leakage etc., radioactive beams, neutron{minus}rich nuclei, neutron{minus}induced and spontaneous fission, ternary fission phenomena, angular momentum, parity and time{minus}reversal phenomena, and nuclear fission at higher excitation energy. Modern spectroscopic tools for gamma spectroscopy as applied to fission were also discussed. There were 53 papers presented at the conference,out of which 3 have been abstracted for the Energy,Science and Technology database.(AIP)

  7. 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.

  8. 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.

  9. Fission and Nuclear Liquid-Gas Phase Transition

    E-print Network

    E. A. Cherepanov; V. A. Karnaukhov

    2007-03-30

    The temperature dependence of the liquid-drop fission barrier is considered, the critical temperature for the liquid-gas phase transition in nuclear matter being a parameter. Experimental and calculated data on the fission probability are compared for highly excited $^{188}$Os. The calculations have been made in the framework of the statistical model. It is concluded that the critical temperature for the nuclear liquid--gas phase transition is higher than 16 MeV.

  10. 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.

  11. Energy production using fission fragment rockets

    SciTech Connect

    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 as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs.

  12. 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 metal ions, Mg and Ca, in the ceramic host phases. The immobilization of rear earth (lanthanide series) fission products in these ceramic host phases will also be studied this year. Cerium oxide is chosen to represent the rear earth fission product for substitution studies in spinel, perovskite and zirconolite ceramic hosts. Cerium has +3 and +4 oxidation states and it can replace some of the trivalent or tetravalent host ions to produce the substitution ceramics such as MgAl2-xCexO4, CaTi1-xCexO3, CaZr1-xCexTi2O7 and CaZrTi2-xCexO7. X-ray diffraction analysis will be used to compare the crystalline structures of the pure ceramic hosts and the substitution phases. SEM-EDX analysis will be used to study the Ce distribution in the ceramic host phases. The range of cerium doping is planned to reach the full substitution of the trivalent or tetravalent ions, Al, Ti and Zr, in the ceramic host phases.

  13. Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

    A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

  14. 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].

  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. (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.

  17. Large-scale fission product containment tests

    NASA Astrophysics Data System (ADS)

    Hilliard, R. K.; Postma, A. K.

    1980-11-01

    The Containment Systems Experiment (CSE) program is reviewed, with emphasis on the inherent processes that remove fission products from containment atmospheres and reduce their leakage to the environment. The CSE containment vessel was sized to represent a 1/5 linear scale model of a typical 1000 MW(e) PWR. Nineteen tests were performed in a stream-air atmosphere simulating post-LOCA conditions. In eight tests containment sprays were operated, in five tests a recirculating filter-adsorber loop was operated, and in six tests only natural, passive processes occurred. Sprays were the most effective in removing airborne iodine and particulate aerosols, followed by the filter loop. Although not as effective as the engineered safety features, natural processes of diffusion to surfaces, reaction with paint, gravity settling, and removal in leak paths are shown to be significant.

  18. Fission product behavior during the PBF (Power Burst Facility) Severe Fuel Damage Test 1-1

    SciTech Connect

    Hartwell, J K; Petti, D A; Hagrman, D L; Jensen, S M; Cronenberg, A W

    1987-05-01

    In response to the accident at Three Mile Island Unit 2 (TMI-2), the United States Nuclear Regulatory Commission (USNRC) initiated a series of Severe Fuel Damage tests that were performed in the Power Burst Facility at the Idaho National Engineering Laboratory to obtain data necessary to understand (a) fission product release, transport, and deposition; (b) hydrogen generation; and (c) fuel/cladding material behavior during degraded core accidents. Data are presented about fission product behavior noted during the second experiment of this series, the Severe Fuel Damage Test 1-1, with an in-depth analysis of the fission product release, transport, and deposition phenomena that were observed. Real-time release and transport data of certain fission products were obtained from on-line gamma spectroscopy measurements. Liquid and gas effluent grab samples were collected at selected periods during the test transient. Additional information was obtained from steamline deposition analysis. From these and other data, fission product release rates and total release fractions are estimated and compared with predicted release behavior using current models. Fission product distributions and a mass balance are also summarized, and certain probable chemical forms are predicted for iodine, cesium, and tellurium. An in-depth evaluation of phenomena affecting the behavior of the high-volatility fission products - xenon, krypton, iodine, cesium, and tellurium - is presented. Analysis indicates that volatile release from fuel is strongly influenced by parameters other than fuel temperature. Fission product behavior during transport through the Power Burst Facility effluent line to the fission product monitoring system is assessed. Tellurium release behavior is also examined relatve to the extent of Zircaloy cladding oxidation. 81 fig., 53 tabs.

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

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

    Transmission electron microscopy characterization of irradiated U-7 wt%Mo dispersion fuel were performed on various U-Mo fuel 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-7 wt%Mo body-centered cubic structure. At a fission density between 3.0 and 4.5 × 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 ?4.5 × 1021 fiss/cm3, the U-7 wt%Mo microstructure starts to undergo grain subdivision and can no longer support the ordered bubble superlattice. The sub-divided fuel grains are less than 500 nm in diameter with what appears to be micron-size fission-gas bubbles present on the grain boundaries. Solid fission products typically 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.

  20. 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.

  1. 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.

  2. 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.

  3. Development of fission gas swelling and release models for metallic nuclear fuels

    E-print Network

    Andrews, Nathan Christopher

    2012-01-01

    Fuel swelling and fission gas generation for fast reactor fuels are of high importance since they are among the main limiting factors in the development of metallic fast reactor fuel. Five new fission gas and swelling ...

  4. 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: (?,xn), (n,?), (?,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.

  5. 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.

  6. 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.

  7. 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.

  8. The behavior of the fission products, as they are released from fission events during nuclear reaction, plays an important role in nuclear fuel performance. Fission product release can occur through grain

    E-print Network

    The behavior of the fission products, as they are released from fission events during nuclear reaction, plays an important role in nuclear fuel performance. Fission product release can occur through grain boundary (GB) at low burnups ; therefore, this study simulates the mass transport of fission gases

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

    NASA Astrophysics Data System (ADS)

    Padgett, Stephen; Wang, Tzu-Fang

    2015-01-01

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

  10. Mechanistic modelling of urania fuel evolution and fission product migration during irradiation and heating

    NASA Astrophysics Data System (ADS)

    Veshchunov, M. S.; Dubourg, R.; Ozrin, V. D.; Shestak, V. E.; Tarasov, V. I.

    2007-05-01

    The models of the mechanistic code MFPR (Module for Fission Product Release) developed by IBRAE in collaboration with IRSN are described briefly in the first part of the paper. The influence of microscopic defects in the UO2 crystal structure on fission-gas transport out of grains and release from fuel pellets is described. These defects include point defects such as vacancies, interstitials and fission atoms, and extended defects such as bubbles, pores and dislocations. The mechanistic description of chemically active elements behaviour (fission-induced) is based on complex association of diffusion-vaporisation mechanism involving multi-phase and multi-component thermo-chemical equilibrium at the grain boundary with accurate calculation of fuel oxidation. In the second part, results of the code applications are given to different situations: normal LWR reactor operation, high temperature annealing, loss of coolant accident (LOCA) and severe accidents conditions.

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

    SciTech Connect

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

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

  12. Corrosion and fission products in primary systems of liquid metal cooled reactors in the USA

    SciTech Connect

    Brehm, W.F.; Colburn, R.P.; Maffei, H.P.; Stinson, W.P.; Bunch, W.L.; Bechtold, R.A.; Olson, W.H.

    1987-01-01

    This paper presents a summary of the work in the USA to support the measurement and control of radionuclides in primary systems of liquid metal cooled reactors. The efforts to characterize and control the ingress of radioactive corrosion and fission products, fuel particles, and radioactivity in gas systems have been quite successful in the USA.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2010-12-01

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

  19. Exotic radioactive beams production using fusion-fission reactions

    NASA Astrophysics Data System (ADS)

    Tarasov, Oleg; Amthor, A. M.; Bazin, D.; Mittig, W.; Morrissey, D. J.; Pereira, J.; Sherrill, B. M.; Villari, A. C. C.; Delaune, O.; Farget, F.; Bastin, B.; Caceres, L.; Kamalou, O.; Saint-Laurent, M. G.; Savajols, H.; Stodel, C.; Thomas, J. C.; Blank, B.; Grevy, S.; Lukyanov, S. M.; Perrot, L.

    2014-09-01

    Fusion-Fission reaction products produced by a 238U beam at 24 MeV/u on Be and C targets were measured in inverse kinematics by use of the LISE fragment separator. The identification of fragments was done using the dE-TKE-Brho-ToF method. Germanium gamma-detectors were placed in the focal plane near the Si stopping telescope to provide an independent verification of the isotope identification via isomer tagging. The experiment demonstrated excellent resolution, in Z, A, and q. The results demonstrate that a fragment separator can be used to produce radioactive beams using fusion-fission reactions in inverse kinematics, and further that in-flight fusion-fission can become a useful production method to identify new neutron-rich isotopes, investigate their properties and study production mechanisms. Mass, atomic number and charge-state distributions are reported for the two reactions. The comparison of the experimental atomic-number and mass distributions combined with the analysis of the isotopic-distributions properties show that between the 9Be and the 12C target, the reaction mechanism changes substantially, evolving from a complete fusion-fission reaction to incomplete fusion or fast fission.

  20. 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.

  1. Production of intense beams of fission fragments at MAFF

    NASA Astrophysics Data System (ADS)

    Bongers, H.; Gross, M.; Habs, D.; Maier, H. J.; Sieber, T.; Thirolf, P. G.; Kester, O.; Köster, U.; Faestermann, T.; von Egidy, T.

    2002-02-01

    At the new Munich high flux reactor FRM-II, the Munich Accelerator for Fission Fragments (MAFF) is under development. The main objective will be the production and study of new very heavy elements (Z>100). To obtain this goal, intense beams of neutron rich isotopes (70fission is considered the most suitable method to produce these isotopes due to the large fission cross section and high thermal neutron flux (>1014n/s*cm2) available at the new reactor. The target ion source design is based on the ANUBIS source at OSIRIS in Studvik, optimized for very high neutron fluxes. Using 1 g of 235U diluted in a graphite target, intensities of several 1011 ions/s for 91Kr, 132Sn, or 144Cs, e.g., are expected after mass separation. These singly charged ions will be charge bred in an electron cyclotron resonance ion source and then injected into the MAFF-LINAC to reach the energies at the Coulomb barrier. The production of intense ion beams of neutron rich isotopes by thermal neutron induced fission, the development of the target ion source, and the development of the fission target will be presented.

  2. Reactivity effects of fission product decay in PWRs

    SciTech Connect

    Aragones, J.M.; Ahnert, C.

    1988-01-01

    The purpose of the work reported in this paper is to analyze the effects of fission product chains with radioactive decay on the reactivity in pressurized water reactor (PWR) cores, calculating their accumulation and absorption rates along fuel burnup at continuous operation and after shutdown periods extending from 1 day to a few months. The authors PWR version of the WIMS-D4 code is first used to obtain the individual number densities, absorption rates, and averaged cross sections for every nuclide of the fission product chains with significant decay rates, as a function of fuel burnup at continuous irradiation. Next, by an auxiliary ad hoc code, these data, have been processed together with the required one for fissile nuclides and boron, also taken from WIMS at each burnup step, to calculate the average or effective values relevant for the analysis and the decay and change in overall absorption after several shutdown times. (1) The reactivity effect of fission product decay changes significantly with the shutdown time. The maximum absorption increase by decay is reached in /approx/ 10 days' shutdown. (2) The dependence with fuel type, enrichment, and burnup is slight, but the change with previous power density is nearly linear, which might be significant after coast-down in previous cycles. (3) For long shutdown periods, the overall reactivity effect of decay in the three fission product chains considered is much less than if only the samarium peak due to /sup 149/Nd is considered.

  3. 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.

  4. Fission product release from highly irradiated LWR fuel

    SciTech Connect

    Lorenz, R.A.; Collins, J.L.; Malinauskas, A.P.; Kirkland, O.L.; Towns, R.L.

    1980-02-01

    A series of experiments was conducted with highly irradiated light-water reactor fuel rod segments to investigate fission products released in steam in the temperature range 500 to 1200/sup 0/C. (Two additional release tests were conducted in dry air.) The primary objectives were to quantify and characterize fission product release under conditions postulated for a spent-fuel transportation accident and for a successfully terminated loss-of-coolant accident (LOCA). In simulated, controlled LOCA-type tests, release at the time of rupture proved to be more significant than the diffusional release that followed. Comparison of the release data for the dry-air tests with the release data of similarly conducted tests in steam indicated significant increases in the releases of iodine, ruthenium, and cesium in air. Various parameters that affect fission product release are discussed, and experimental observations and analysis of the chemical behavior of releasable fission products in inert, steam, and dry-air atmospheres are examined.

  5. (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.

  6. 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.

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

    SciTech Connect

    Blaise Collin

    2014-09-01

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

  8. 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.

  9. Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

    To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced ?-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

  10. 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)

  11. Analysis of the Transmutation of Long Lived Fission Products Using an Accelerator 

    E-print Network

    Hearne, Jason A

    2015-04-07

    The aim of this study is to analyze the transmutation of Long Lived Fission Products (LLFPs) using an accelerator based system. The seven LLFP’s are investigated based upon their decay mechanics, yields from fission, and neutron absorption cross...

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

    DOEpatents

    Herrmann, Steven Douglas

    2014-05-27

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

  13. Fission product precipitates in irradiated uranium carbonitride fuel

    NASA Astrophysics Data System (ADS)

    Kleykamp, H.

    2002-02-01

    Austenitic steel-cladded uranium carbonitride fuel pins were irradiated in the BR2 up to 6.4% burnup. A cross-section of the pin RV 24 with the fuel composition UC 0.86N 0.09O 0.05 was prepared for X-ray microanalysis of the fission product precipitates. Rare-earth oxide and U(Mo,Tc)C 2 phases were observed in the whole fuel region. Bright phases present in annular rings of the outer fuel zone were identified as U 2(Tc, Ru, Rh)C 2. Alkaline-earth oxide and U-Pd-Ni phases were shown in the fuel-cladding gap. The rare-earth and alkaline-earth fission products extracted the oxygen from the fuel matrix which became nearly oxygen free. The formation of nitrides could not be detected.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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. This combination consumes about 20% of the thorium initially loaded in the hybrid reactor ({approx}200 GWd/tHM), partially during hybrid operation, but mostly during operation in the critical reactor. The plant support ratio is low compared to the one attainable using continuous fuel chemical reprocessing, which can yield a plant support ratio of about 20, but the resulting fuel cycle offers better proliferation resistance as fissile material is never separated from the other fuel components.

  19. 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.

  20. 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.

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

    SciTech Connect

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

    2014-10-01

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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Methodology and application of the WIMS-D4M fission product data

    SciTech Connect

    Mo, S.C.

    1995-02-01

    The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the {sup 135}Xe and {sup 149}Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements.

  8. Measurement of fission product yields in the quasi-mono-energetic neutron-induced fission of 238U

    NASA Astrophysics Data System (ADS)

    Naik, H.; Mukerji, Sadhana; Crasta, Rita; Suryanarayana, S. V.; Sharma, S. C.; Goswami, A.

    2015-09-01

    The cumulative yields of various fission products in the 6.35, 8.53, 9.35 and 12.52 MeV quasi-mono-energetic neutron induced fission of 238U have been determined by using the off-line ?-ray spectrometric technique. The mass chain yields were obtained from the fission product yields by using charge distribution correction. From the mass yield data, the peak-to-valley (P/V) ratio, the average value of light mass (), heavy mass () and thus the average number of neutrons (expt) were obtained in the 238U(n, f) reaction for the above mentioned four neutron energies. The present and literature data in the 238U(n, f) and 238U (?, f) reactions at various energies were compared to arrive at the following conclusions. (i) The yields of fission products for A = 133- 134, A = 138- 140, and A = 143- 144 and their complementary products in the 238U(n, f) and 238U(?, f) reactions are higher than other fission products, which has been explained from the point of even-odd effect and standard I and standard II asymmetric mode of fission. (ii) The yields of symmetric products increase and thus the peak-to-valley (P/V) ratios decrease with excitation energy, whereas the expt values increase with excitation energy. (iii) The variation of , and expt values with excitation energy behave differently in between 238U(n, f) and 238U(?, f) reactions, which may be due to the different types of reaction mechanism for the neutron and photon with 238U.

  9. 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.

  10. 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).

  11. 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.

  12. 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.

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

    DOEpatents

    Vandegrift, George F. (Bolingbrook, IL); Vissers, Donald R. (Naperville, IL); Marshall, Simon L. (Woodridge, IL); Varma, Ravi (Hinsdale, IL)

    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.

  14. 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.

  15. 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)

  16. 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.

  17. Structural stability and fission product behaviour in U3Si

    NASA Astrophysics Data System (ADS)

    Middleburgh, S. C.; Burr, P. A.; King, D. J. M.; Edwards, L.; Lumpkin, G. R.; Grimes, R. W.

    2015-11-01

    The crystalline and amorphous structures of U3Si have been investigated using density functional theory techniques for the first time. The effects of disorder and the impact of fission products has been separated to understand the swelling characteristics of U3Si in both crystalline and amorphous U3Si. Initially, the stability of the three experimentally observed polymorphs of U3Si were explored. Subsequently, we modelled the amorphous U3Si system and conclude that initial increase in volume observed experimentally at low temperature corresponds well with the volume change that occurs with the observed amorphisation of the material. The solubility of Xe and Zr into both the crystalline and amorphous systems was subsequently investigated.

  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. Neutron Cross Section Covariances for Structural Materials and Fission Products

    SciTech Connect

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

    2011-12-01

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

  20. Design of a Gas Delivery System for use with a Fission Time Projection Chamber

    NASA Astrophysics Data System (ADS)

    Snyder, Lucas; Greife, Uwe

    2009-10-01

    Developing advanced nuclear reactors and waste recycling techniques requires an improvement in many basic nuclear physics measurements. To address part of this need a Time Projection Chamber is being constructed to measure fission cross sections to a higher precision than traditional fission chambers. This talk will discuss how a Time Projection Chamber works and the details of a gas delivery system being constructed for use with it.

  1. Fission products in nuclear fuels pellets can affect fuel performance as they change the fuel chemistry and structure. The behavior of the fission products and their release mechanisms are important to the operation

    E-print Network

    Fission products in nuclear fuels pellets can affect fuel performance as they change the fuel chemistry and structure. The behavior of the fission products and their release mechanisms are important to the operation of a power reactor. Research has shown that fission product release can occur through grain

  2. 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.

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

    NASA Astrophysics Data System (ADS)

    Barber, Duncan Henry

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

  4. Secretory production of ricinoleic acid in fission yeast Schizosaccharomyces pombe.

    PubMed

    Yazawa, Hisashi; Kumagai, Hiromichi; Uemura, Hiroshi

    2013-10-01

    We have succeeded to produce a high content of ricinoleic acid (RA), a hydroxylated fatty acid with great values as a petrochemical replacement, in fission yeast Schizosaccharomyces pombe by introducing Claviceps purpurea oleate ?12-hydroxylase gene (CpFAH12). Although the production was toxic to S. pombe cells, we solved the problem by identifying plg7, encoding phospholipase A2, as a multicopy suppressor. Characterization of the RA-tolerant strains suggested that the removal of RA moieties from phospholipids would be the suppression mechanism by plg7. In this study, we extended our analysis and report our new discovery that the overexpression of plg7 enabled cells to secrete free RA into culture media. When the FAH12 integrant in the absence of the overexpressed plg7 was grown at 20 °C for 11 days, the amount of intracellular RA reached 200.1 ?g/ml of culture and only 69.3 ?g/ml of RA was detected in culture media. On the other hand, the FAH12 integrant harboring the plg7 multicopy plasmid secreted RA in the media (184.5 ?g/ml) without decreasing the amount in the cells, i.e., a significantly higher total secretion and a lead to making RA by its secretory production in S. pombe. PMID:23820557

  5. 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)

  6. 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.

  7. Results of fission product release from intermediate-scale MCCI (molten core-concrete interaction) tests

    SciTech Connect

    Spencer, B.W.; Thompson, D.H.; Fink, J.K.; Gunther, W.H.; Sehgal, B.R.

    1988-01-01

    A program of reactor-material molten core-concrete interaction (MCCI) tests and related analyses are under way at Argonne National Laboratory under sponsorship of the Electric Power Research Institute (EPRI). The particular objective of these tests is to provide data pertaining to the release of nonvolatile fission products such as La, Ba, and Sr, plus other aerosol materials, from the coupled thermal-hydraulic and chemical processes of the MCCI. The first stages of the program involving small and intermediate-scale tests have been completed. Three small-scale tests (/approximately/5 kg corium) and nine intermediate-scale tests (/approximately/30 kg corium) were performed between September 1985 and September 1987. Real reactor materials were used in these tests. Sustained internal heat generation at nominally 1 kW per kg of melt was provided by direct electrical heating of the corium mixture. MCCI tests were performed with both fully and partially oxidized corium mixtures that contained a variety of nonradioactive materials such as La/sub 2/O/sub 3/, BaO, and SrO to represent fission products. Both limestone/common sand and basaltic concrete basemats were used. The system was instrumented for characterization of the thermal hydraulic, chemical, gas release, and aerosol release processes.

  8. 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)

  9. 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.

  10. 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.

  11. Sorption of some fission products and actinides in concrete systems

    SciTech Connect

    Hoeglund, S.; Eliasson, L.; Allard, B.; Andersson, K.; Torstenfelt, B.

    1986-01-01

    The sorption of some actinides (Th, U, Np, Pu and Am) and fission products (I, Cs) was measured on two types of Standard Portland cements as well as on samples from old (70 years) hydro power dam constructions using a batch technique. Pore water compositions were analyzed, and artificial pore water solutions were used as aqueous phases in the experiments. Measurements were also performed on five other concrete types (not reported in detail in this paper) to illustrate the effects of the cement matrix composition on the sorption behavior of the radionuclides. The sorption of actinides in the trivalent (americium), tetravalent (thorium) pentavalent (neptunium) and hexavalent (uranium) states was high in all the studied concrete systems. Generally, the sorption of cesium was low due to the low exchange capacity of the cement and the high concentration of competing cations in the pore waters. The sorption of iodine was much higher than in most silicate minerals of geologic origin. The differences between the various concrete systems were generally minor in terms of their sorbing capacities.

  12. Fission gas release and swelling in uranium-plutonium mixed nitride fuels

    NASA Astrophysics Data System (ADS)

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

    2004-05-01

    Two uranium-plutonium mixed nitride, (U,Pu)N, fuel pins with different He-gap width were irradiated at a linear heating rate 75 kW/m to 4.3% FIMA in the experimental fast reactor JOYO, and nondestructive and destructive post irradiation examinations were carried out. Fission gas release rates were about 3.3% and 5.2%, and swelling rates were about 1.8% and 1.6%/% FIMA. From the radial distributions of Xe concentration measured by EPMA, it was determined that approximately 80% and 15% of fission gases were retained in the intragranular region and in the fission gas bubbles, respectively. Deformation of the fuel cladding differed between the two tested fuel pins. A uniform diameter increase was observed in the small gap fuel pin, while ovalities, which seemed to be caused by relocation of the fuel fragments, were found in the large gap one.

  13. 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.

  14. 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

  15. 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.

  16. 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 ?}? 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.

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

    SciTech Connect

    Alexandra Navrotsky

    2010-07-30

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

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

    NASA Technical Reports Server (NTRS)

    1973-01-01

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

  19. Towards a multiscale approach for assessing fission product behaviour in UN

    NASA Astrophysics Data System (ADS)

    Klipfel, M.; Di Marcello, V.; Schubert, A.; van de Laar, J.; Van Uffelen, P.

    2013-11-01

    Ab initio modelling of fission products (i.e. Nb, Y, Gd, Nd, Zr, Sm, Eu, Ce, Ba, Mo, Sr, Rh, Pd, and Ru) in uranium nitride is carried out by assessing the incorporation, along with their contributions to local swelling of the fuel matrix. Fission products (FP's) in UN have shown to be preferably accommodated at U vacancies in bound [1 0 0]-Schottky defects, nevertheless, similar incorporation energies were found at a single U vacancy. From the investigated incorporation and migration mechanism, we found that FP's in UN predominately migrate along U-U vacancies, since the incorporation energies for all FP are lowest at single U vacancy or at the U vacancy in a Schottky defect. The energy required to induce a migration of a volatile FP from an N vacancy to U vacancy is about 4-5.5 eV. The local volume changes caused by the fission-product substitution have been assessed by means of DFT and combined with the fission-product concentrations obtained by means of neutron calculations (SCALE) to predict fission product swelling in UN. The linear swelling of nitride fuel resulting from these calculations, and the assumption that fission products do not interact and form secondary phases, leads to a reasonable estimation for the swelling rate as a function of burn-up (or time) when compared with empirical correlations in the open literature.

  20. 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.

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

    E-print Network

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

    2008-01-29

    Critical temperature Tc for the nuclear liquid-gas phase transition is stimated both from the multifragmentation and fission data. In the first case,the critical temperature is obtained by analysis of the IMF yields in p(8.1 GeV)+Au collisions within the statistical model of multifragmentation (SMM). In the second case, the experimental fission probability for excited 188Os is compared with the calculated one with Tc as a free parameter. It is concluded for both cases that the critical temperature is higher than 16 MeV.

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

    DOEpatents

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

    2007-10-02

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

  3. High resolution axial ionization chamber for fission products

    NASA Astrophysics Data System (ADS)

    Oed, A.; Geltenbort, P.; Gönnenwein, F.; Manning, T.; Souque, D.

    1983-02-01

    The construction of an ionization chamber with the electrical field parallel to the particle beam (axial chamber) is described. The energy resolution achieved for typical light and heavy mass fragments from the thermal neutron induced fission of 235U is 385 keV and 510 keV, respectively.

  4. 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.

  5. 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.

  6. 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.

  7. Glasses for Immobilizing Lanthanide, Alkali, and Alkali-earth Fission Products

    SciTech Connect

    Crum, Jarrod V.; Vienna, John D.

    2009-08-03

    A series of glasses were formulated for the immobilization of a potential waste stream from commercial nuclear fuel reprocessing, the combined lanthanide (LN), alkali, and alkaline earth (Cs/Sr) fission products. These glasses were formulated to meet repository disposal requirements while being processable in a cold-crucible melter. The glasses were fabricated and tested for product consistency test response, phase characterization, density, and glass transition temperature. The results suggest that the combined fission product waste forms are likely to meet repository requirements and generate less glass than if individual streams were vitrified.

  8. Transient fission-gas behavior in uranium nitride fuel under proposed space applications. Doctoral thesis

    SciTech Connect

    Deforest, D.L.

    1991-12-01

    In order to investigate whether fission gas swelling and release would be significant factors in a space based nuclear reactor operating under the Strategic Defense Initiative (SDI) program, the finite element program REDSTONE (Routine For Evaluating Dynamic Swelling in Transient Operational Nuclear Environments) was developed to model the 1-D, spherical geometry diffusion equations describing transient fission gas behavior in a single uranium nitride fuel grain. The equations characterized individual bubbles, rather than bubble groupings. This limits calculations to those scenarios where low temperatures, low burnups, or both were present. Instabilities in the bubble radii calculations forced the implementation of additional constraints limiting the bubble sizes to minimum and maximum (equilibrium) radii. The validity of REDSTONE calculations were checked against analytical solutions for internal consistency and against experimental studies for agreement with swelling and release results.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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 release path. This exercise has identified a host of material properties, thermofluid states, and physics models that must be collected, defined, and understood to evaluate this attenuation factor. The assembled PIRT table underwent two iterations with extensive reorganization between meetings. Generally, convergence was obtained on most issues, but different approaches to the specific physics and transport paths shade the answers accordingly. The reader should be cautioned that merely selecting phenomena based on high importance and low knowledge may not capture the true uncertainty of the situation. This is because a transport path is composed of several serial linkages, each with its own uncertainty. The propagation of a chain of modest uncertainties can lead to a very large uncertainty at the end of a long path, resulting in a situation that is of little regulatory guidance.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

    Prompt gamma-ray 4-fold data were built to collect 2×1011 ? -? -? -? quadruple- and higher-fold ? -coincidence events from the spontaneous fission of 252Cf with Gammasphere detector arrays. The nuclei 106Nb, 115Pd, 142La, 145,146Ba, 152Ce and Gd have been studied with these data. By using the new 4-fold data, we confirmed several weak tentative transitions in 106Nb, 142La, 145,146Ba, 148Ce which were observed previously from the ? -? -? triple cube. Some new transitions in 106Nb, 142La were identified by our new 4-fold data. Cascades in 145,146Ba are much clearer in four-fold data than the previous triple coincidence data. We will continue to study other nuclei by our 4-fold data with lower background than the previous triple cube.

  19. A 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.

  20. 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…

  1. 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.

  2. 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...

  3. 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

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of {approx}0 at% U-235 (LEU) or a fission density of {approx}3 x 10{sup 21} fissions/cm{sup 3}. Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed that the site of first bubble appearance is the grain boundary. Analysis using a simple diffusion model showed that, although the difference in the Mo-content between the grain boundary and grain interior region decreased with burnup, a complete convergence in the Mo-content was not reached at the end of the test for all RERTR tests. A total of 13 plates from RERTR-1, 2, 3 and 5 tests with different as-fabrication conditions and irradiation conditions were included for gas bubble analyses. Among them, two plates contained powders {gamma}-annealed at {approx}800 C for {approx}100 hours. Most of the plates were fabricated with as-atomized powders except for two as-machined powder plates. The Mo contents were 6, 7 and 10wt%. The irradiation temperature was in the range 70-190 C and the fission rate was in the range 2.4 x 10{sup 14} - 7 x 10{sup 14} f/cm{sup 3}-s. Bubble size for both of the {gamma}-annealed powder plates is smaller than the as-atomized powder plates. The bubble size for the as-atomized powder plates increases as a function of burnup and the bubble growth rate shows signs of slowing at burnups higher than {approx}40 at% U-235 (LEU). The bubble-size distribution for all plates is a quasi-normal, with the average bubble size ranging 0.14-0.18 {micro}m. Although there are considerable errors, after an initial incubation period the average bubble size increases with fission density and shows saturation at high fission density. Bubble population (density) per unit grain boundary length was measured. The {gamma}-annealed powder plates have a higher bubble density per unit grain boundary length than the as-atomized powder plates. The measured bubble number densities per unit grain boundary length for as-atomized powder plates are approximately constant with respect to burnup. Bubble density per unit cross section area was calculated using the density per unit grain boundary length data. The grains were modeled as tetrakaidecahedrons. Direct measurements for some plates were also performed and compared with the calculated quantities. Bubble density per unit

  4. 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

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

  5. Fission gas transport and its interaction with irradiation induced defects in lanthanum doped ceria

    NASA Astrophysics Data System (ADS)

    Yun, Di

    Combined experimental and modeling efforts have been extremely productive in understanding irradiation-induced displacement damage in metal and metal alloy systems. In order to help understand the fundamental mechanisms of irradiation-induced defect formation and evolution in nuclear fuel, similar combined modeling and experimental efforts have been carried out. Ceria (CeO2) was selected as a surrogate material for Uranium Dioxide (UO2) due to its many similar properties. Lanthanum (La) was chosen as a dopant in CeO 2 to investigate the effect of impurities in a controlled manner. The presence of La in the CeO2 lattice introduces a predictable initial concentration of oxygen vacancies, making it possible to characterize hypo-stoichiometric effects in CeO2. The influence of two La concentrations, 5% and 25%, were examined. Radiation damage was induced using low energy ion implantations and high energy ion irradiation experiments, where the ion beam energy was selected for high displacement damage levels and/or high levels of implanted Xe or Kr. A combination of in situ TEM (Transmission Electron Microscopy) and ex situ TEM experiments were used to study the evolution of defect clusters and the influence of two common fission products, Xe and Kr. The irradiations were performed on thin film, single crystal materials so that the material composition and crystallinity could be directly controlled. The irradiation damage caused the formation of complex microstructures with dislocation loops, voids or bubbles, and dislocation networks at higher doses. The Burgers vectors of the dislocation loops were determined and the loops were found to be mainly [111] type Burgers vector pure edge loops. They have been tentatively identified as interstitial type. La, as an impurity, has revealed a strong defect trapping effect. Various sets of quantitative experimental results were obtained to characterize the dose and temperature effects of irradiation. These results also help to benchmark simulation codes being developed with a kinetic Monte Carlo model. These experimental results include size and size distributions of dislocation loops, voids and gas bubble structures created by irradiation. More importantly, this systematic experimental work has provided key insights into the understanding of the mechanisms of defect evolution in the materials investigated. A model including both defect production and annihilation mechanisms has been proposed to explain the observed defect kinetics in the lower dose regime. A coalescence driven model has been proposed for void/bubble growth in the higher dose regime. Experimental results also revealed that lanthanum trapping has significant influence on the void/bubble growth in the CeO2 lattice. Lattice and kinetic Monte Carlo calculations have provided key insights to the interpretations of experimental results.

  6. 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.

  7. 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 values might be somewhat over-estimated because the strontium retention during irradiation cannot be assessed a priori, which affects the magnitude of the calculated release during safety testing. The diffusivity of strontium in UCO cannot be derived from these heating tests, but it is assumed to be modeled correctly using the IAEA recommended value for kernel diffusivity. • For krypton, there is no reliable release data for compacts heated up at 1600°C, which includes all the compacts containing only intact particles. At 1700 and 1800°C, comparisons show an over-prediction of the release from compacts containing particles with failed SiC by 1 to 1.5 orders of magnitude. The available data from these heating tests do not allow to determine which of the TRISO-coating’s layers diffusivities are under or over-estimated.

  8. 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.

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

    SciTech Connect

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

    2007-04-22

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

  10. Fission product Pd-SiC interaction in irradiated coated particle fuels

    SciTech Connect

    Tiegs, T.N.

    1980-04-01

    Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO/sub 2/ to UC/sub 2/, including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% /sup 235/U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary.

  11. Swelling due to fission products and additives dissolved within the uranium dioxide lattice

    NASA Astrophysics Data System (ADS)

    Middleburgh, S. C.; Grimes, R. W.; Desai, K. H.; Blair, P. R.; Hallstadius, L.; Backman, K.; Van Uffelen, P.

    2012-08-01

    Simulations using empirical inter-atomic potentials have been used to predict the change in volume of the uranium dioxide lattice due to the accommodation of soluble fuel additives and fission products. The incorporation of divalent, trivalent and tetravalent cations are considered. The change in accommodation mechanism for aliovalent cations between UO2 and UO2+x gives rise to markedly different defect volumes. Experimental data is in good agreement with the predictions made in this work, particularly swelling as a function of dopant concentration under different conditions. The predicted defect volumes have been combined to predict the change in lattice volume with burnup (fission product inventory) due to incorporation of these soluble species, which agrees well with swelling data from irradiated fuel.

  12. 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.

  13. Fission-Product Separation Based on Room-Temperature Ionic Liquids

    SciTech Connect

    Luo, Huimin; Hussey, Charles L.

    2005-09-30

    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 extraction systems based on ionic liquids; (c) to develop efficient processes to recycle ionic liquids and crown ethers; and (d) to investigate chemical stabilities of ionic liquids under strong acid, strong base, and high-level-radiation conditions.

  14. Fission-Product Separation Based on Room-Temperature Ionic Liquids

    SciTech Connect

    Luo, Huimin

    2006-11-15

    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 extraction systems based on ionic liquids; (c) to develop efficient processes to recycle ionic liquids and crown ethers; and (d) to investigate chemical stabilities of ionic liquids under strong acid, strong base, and high-level-radiation conditions.

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

    PubMed

    Abrecht, David G; Schwantes, Jon M

    2015-03-01

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

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

    SciTech Connect

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

    1988-01-01

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

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

    SciTech Connect

    Joshua R. Versey

    2013-08-01

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

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

    SciTech Connect

    Kwong, S.; Small, J.

    2007-07-01

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

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

    SciTech Connect

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

    1988-01-01

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

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

    SciTech Connect

    Abrecht, David G.; Schwantes, Jon M.

    2015-03-03

    This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln??=-? (?G_rxn^° (T_C ))/(RT_C )+? were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ?G_rxn^° (T_C ). These models allowed an estimate of the upper bound for the reactor temperatures of T_C between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

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

    SciTech Connect

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

    2014-07-07

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

  2. Disposition of plutonium-239 via production of fission molybdenum-99.

    PubMed

    Mushtaq, A

    2011-04-01

    A heritage of physical consequences of the U.S.-Soviet arms race has accumulated, the weapons-grade plutonium (WPu), which will become excess as a result of the dismantlement of the nuclear weapons under the arms reduction agreements. Disposition of Pu has been proposed by mixing WPu with high-level radioactive waste with subsequent vitrification into large, highly radioactive glass logs or fabrication into mixed oxide fuel with subsequent irradiation in existing light water reactors. A potential option may be the production of medical isotope molybdenum-99 by using Pu-239 targets. PMID:21256759

  3. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    NASA Astrophysics Data System (ADS)

    Wang, Xin-Hua; Guo, Hai-Ping; Mou, Yun-Feng; Zheng, Pu; Liu, Rong; Yang, Xiao-Fei; Yang, Jian

    2013-05-01

    A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D+ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three-dimensional Monte Carlo code MCNP5 and the ENDF/B-VI data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(?, ?) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.

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

    SciTech Connect

    James Stubbins

    2012-12-19

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

  5. Yrast states of neutron-rich N=83 nuclei from fission product {gamma}-ray studies

    SciTech Connect

    Bhattacharyya, P.; Zhang, C.T.; Fornal, B.; Daly, P.J.; Grabowski, Z.W.; Ahmad, I.; Lauritsen, T.; Morss, L.R.; Phillips, W.R.; Durell, J.L.; Leddy, M.J.; Smith, A.G.; Urban, W.; Varley, B.J.; Schulz, N.; Lubkiewicz, E.; Bentaleb, M.; Blomqvist, J.

    1997-11-01

    Prompt {gamma}-ray cascades in N=83 fission product nuclei near {sup 132}Sn have been studied at Eurogam II using a {sup 248}Cm source. Cross coincidences observed between {gamma} rays from complementary light and heavy fission fragments were vital for isotopic assignments. Yrast states in the N=83 isotones {sup 134}Sb, {sup 135}Te, and {sup 136}I are reported. The interpretation of the level schemes is based mainly on results of shell model calculations using empirical proton-proton interaction energies from {sup 134}Te, and proton-neutron interactions estimated from the well-known {sup 210}Bi level spectrum. {copyright} {ital 1997} {ital The American Physical Society}

  6. 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.

  7. 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)

  8. Development of a gas-jet coupled ISOL facility with a /sup 252/Cf spontaneous fission source

    E-print Network

    Greenwood, R C; Novick, V J

    1981-01-01

    A mass separator at the INEL has been successfully coupled on-line to a source of /sup 252/Cf fission products via a He-gas jet transport arrangement using solid aerosols of NaCl as activity carriers. Initial tests of the ISOL system on-line to an approximately 7 mu g /sup 252 /Cf source are conducted using gamma-ray spectroscopic measurements of the separated /sup 138,139/Cs, /sup 141,142/Ba and /sup 142/La activities. The measured transport efficiencies through the system of approximately 3% and approximately 0.3% for the Cs and Ba isotopes, respectively, are comparable with the results of earlier tests conducted at INEL with a hollow-cathode ion source alone coupled to the He-gas jet transport arrangement. Following these tests, a general survey of the mass-separated activities is conducted with the ISOL system on-line to an approximately 600 mu g source of /sup 252/Cf. Gross beta - gamma activity is measured for samples collected at 73 mass positions. Gamma-ray spectra are measured with a Ge(Li) detector ...

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

    DOEpatents

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

    2009-05-05

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

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

    DOEpatents

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

    2009-01-06

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

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

    DOEpatents

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

    2009-01-27

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

  12. Isomer production ratios and the angular momentum distribution of fission fragments

    NASA Astrophysics Data System (ADS)

    Stetcu, I.; Talou, P.; Kawano, T.; Jandel, M.

    2013-10-01

    Latest generation fission experiments provide an excellent testing ground for theoretical models. In this contribution we compare the measurements for 235U(nth,f), obtained with the Detector for Advanced Neutron Capture Experiments (DANCE) calorimeter at Los Alamos Neutron Science Center (LANSCE), with our full-scale simulation of the primary fragment de-excitation, using the recently developed cgmf code, based on a Monte Carlo implementation of the Hauser-Feshbach theoretical model. We compute the isomer ratios as a function of the initial angular momentum of the fission fragments, for which no direct information exists. Comparison with the available experimental data allows us to determine the initial spin distribution. We also study the dependence of the isomer ratio on the knowledge of the low-lying discrete spectrum input for nuclear fission reactions, finding a high degree of sensitivity. Finally, in the same Hauser-Feshbach approach, we calculate the isomer production ratio for thermal neutron capture on stable isotopes, where the initial conditions (spin, excitation energy, etc.) are well understood. We find that with the current parameters involved in Hauser-Feshbach calculations, we obtain up to a factor of 2 deviation from the measured isomer ratios.

  13. Shale gas production: potential versus actual greenhouse gas emissions*

    E-print Network

    Shale gas production: potential versus actual greenhouse gas emissions* Francis O Environ. Res. Lett. 7 (2012) 044030 (6pp) doi:10.1088/1748-9326/7/4/044030 Shale gas production: potential gas (GHG) emissions from shale gas production and use are controversial. Here we assess the level

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

    NASA Astrophysics Data System (ADS)

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

    2014-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  16. Rapid monitoring of gaseous fission products in BWRs using a portable spectrometer

    SciTech Connect

    Yeh, Wei-Wen; Lee, Cheng-Jong; Chen, Chen-Yi; Chung, Chien

    1996-12-31

    Rapid, quantitative determination of gaseous radionuclides is the most difficult task in the field of environmental monitoring for radiation. Although the identification of each gaseous radionuclide is relatively straightforward using its decayed gamma ray as an index, the quantitative measurement is hampered by the time-consuming sample collection procedures, in particular for the radioactive noble gaseous fission products of krypton and xenon. In this work, a field gamma-ray spectrometer consisting of a high-purity germanium detector, portable multichannel anlayzer, and a notebook computer was used to conduct rapid scanning of radioactive krypton and xenon in the air around a nuclear facility.

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

    SciTech Connect

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

    1997-01-01

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

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

    SciTech Connect

    Jenkins, I.L.

    1984-01-01

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

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

    SciTech Connect

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

    1994-12-31

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

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

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.

    2005-07-01

    A new three objective optimization technique is developed and applied to find the operating conditions for fission reactor heated Closed Cycle Gas Turbine (CCGT) space power systems at which maximum efficiency, minimum radiator area, and minimum total system mass is achieved. Such CCGT space power systems incorporate a nuclear reactor heat source with its radiation shield; the rotating turbo-alternator, consisting of the compressor, turbine and the electric generator (three phase AC alternator); and the heat rejection subsystem, principally the space radiator, which enables the hot gas working fluid, emanating from either the turbine or a regenerative heat exchanger, to be cooled to compressor inlet conditions. Numerical mass models for all major subsystems and components developed during the course of this work are included in this report. The power systems modeled are applicable to future interplanetary missions within the Solar System and planetary surface power plants at mission destinations, such as our Moon, Mars, the Galilean moons (Io, Europa, Ganymede, and Callisto), or Saturn's moon Titan. The detailed governing equations for the thermodynamic processes of the Brayton cycle have been derived and successfully programmed along with the heat transfer processes associated with cycle heat exchangers and the space radiator. System performance and mass results have been validated against a commercially available non-linear optimization code and also against data from existing ground based power plants.

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

    SciTech Connect

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

    2012-05-30

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

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

    SciTech Connect

    Laurec, J.; Adam, A.; Bruyne, T. de; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

    2010-12-15

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

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

    SciTech Connect

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

    1984-01-01

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

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

    SciTech Connect

    Martin, R.C.

    1990-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2009-06-01

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

  7. Investigation of Fission Product Transport into Zeolite-A for Pyroprocessing Waste Minimization

    SciTech Connect

    James R. Allensworth; Michael F. Simpson; Man-Sung Yim; Supathorn Phongikaroon

    2013-02-01

    Methods to improve fission product salt sorption into zeolite-A have been investigated in an effort to reduce waste associated with the electrochemical treatment of spent nuclear fuel. It was demonstrated that individual fission product chloride salts were absorbed by zeolite-A in a solid-state process. As a result, recycling of LiCl-KCl appears feasible via adding a zone-freezing technique to the current treatment process. Ternary salt molten-state experiments showed the limiting kinetics of CsCl and SrCl2 sorption into the zeolite. CsCl sorption occurred rapidly relative to SrCl2 with no observed dependence on zeolite particle size, while SrCl2 sorption was highly dependent on particle size. The application of experimental data to a developed reaction-diffusion-based sorption model yielded diffusivities of 8.04 × 10-6 and 4.04 × 10-7 cm2 /s for CsCl and SrCl2, respectively. Additionally, the chemical reaction term in the developed model was found to be insignificant compared to the diffusion term.

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

    SciTech Connect

    Kleykamp, H.

    1988-03-01

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

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

    SciTech Connect

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

    2012-10-01

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

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

    SciTech Connect

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

    2014-05-01

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

  11. 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).

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

    SciTech Connect

    Wright, R.Q.

    1999-01-01

    In response to concerns raised in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-2, the U.S. Department of Energy (DOE) developed a comprehensive program to help assure that the DOE maintain and enhance its capability to predict the criticality of systems throughout the complex. Tasks developed to implement the response to DNFSB recommendation 93-2 included Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data Task consists of a program of differential measurements at the Oak Ridge Electron Linear Accelerator (ORELA), precise fitting of the differential data with the generalized least-squares fitting code SAMMY to represent the data with resonance parameters using the Reich-Moore formalism along with covariance (uncertainty) information, and the development of complete evaluations for selected nuclides for inclusion in the Evaluated Nuclear Data File (ENDFB). The current ENDF/B library was developed for fast and thermal fission reactors and fusion reactors. Criticality safety practitioners recognize that many situations around the DOE complex are characterized by neutron spectra in the intermediate-energy region, as opposed to the high-energy region for fast reactors and fusion systems and the low-energy region for thermal reactors. Consequently, the Nuclear Data Task focuses primarily on the intermediate-energy region so that upgrades to existing evaluated data will remove deficiencies in the current ENDF/B evaluations. The ORELA allows high-resolution measurements in the intermediate-energy region and the SAMMY fitting code provides high quality resonance parameters in the resolved and unresolved energy range using the sophisticated Reich-Moore (RM) formalism for superior representation of the data in the intermediate energy region. In addition, the SAMMY fitting procedure provides covariance information for the resonance parameters that can be used in subsequent analyses to assess the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature.

  13. Kinetic study of fission product activity released inside containment under loss of coolant transients in a typical MTR system.

    PubMed

    Awan, Saeed E; Mirza, Nasir M; Mirza, Sikander M

    2012-12-01

    Based on continuous release of fission product (FP) activity from fuel to the coolant and then to the containment, a kinetic model is developed for source term after a LOCA in a typical MTR type system. The time dependent source, re-suspension rate, decay of fission products, leakage, deposition on surfaces, and re-circulation of air through filters are employed with a partial prompt source plus a time varying source. Releases of different FP activities are simulated for various release rates. PMID:23041390

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

    SciTech Connect

    Rest, J.

    1995-08-01

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

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

    SciTech Connect

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

    2014-01-01

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

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

    E-print Network

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

    2012-09-13

    In this paper, we study the impact of the inclusion of the recently measured beta decay properties of the $^{102;104;105;106;107}$Tc, $^{105}$Mo, and $^{101}$Nb nuclei in an updated calculation of the antineutrino energy spectra of the four fissible isotopes $^{235, 238}$U, and $^{239,241}$Pu. These actinides are the main contributors to the fission processes in Pressurized Water Reactors. The beta feeding probabilities of the above-mentioned Tc, Mo and Nb isotopes have been found to play a major role in the $\\gamma$ component of the decay heat of $^{239}$Pu, solving a large part of the $\\gamma$ discrepancy in the 4 to 3000\\,s range. They have been measured using the Total Absorption Technique (TAS), avoiding the Pandemonium effect. The calculations are performed using the information available nowadays in the nuclear databases, summing all the contributions of the beta decay branches of the fission products. Our results provide a new prediction of the antineutrino energy spectra of $^{235}$U, $^{239,241}$Pu and in particular of $^{238}$U for which no measurement has been published yet. We conclude that new TAS measurements are mandatory to improve the reliability of the predicted spectra.

  17. Shale gas production: potential versus actual greenhouse gas emissions

    E-print Network

    O’Sullivan, Francis Martin

    Estimates of greenhouse gas (GHG) emissions from shale gas production and use are controversial. Here we assess the level of GHG emissions from shale gas well hydraulic fracturing operations in the United States during ...

  18. Thermodynamics of vaporization of fission products and materials under severe reactor accident conditions: Analysis of molten core/concrete chemistry

    NASA Astrophysics Data System (ADS)

    Cubicciotti, Daniel

    1985-02-01

    Vaporization-condensation processes can generate radioactive aerosols in the event of a core dryout and meltdown accident at a nuclear power station. The time sequence of fission produce vaporization and aerosol formation in relation to processes that can transport them out of the reactor containment is important for assessing their potential biohazard. Thermodynamics of vaporization of fission products and other materials are evaluated for the extreme environmental conditions projected by computer models if a molten core penetrates the reactor vessel and melts into the concrete base. A free energy minimization treatment was used to estimate partial pressures of gases in this many-component, multiphase system. The amounts of fission products and condensable materials vaporized were calculated for a test case involving basalt-aggregate concrete.

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

    NASA Astrophysics Data System (ADS)

    Claisse, Antoine; Van Uffelen, Paul

    2015-11-01

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

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

    SciTech Connect

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

    2007-02-28

    Recent field investigation that collected and characterized vadose zone sediments from beneath inactive liquid disposal facilities at the Hanford 200 Areas show lower than expected concentrations of a long-term risk driver, Tc-99. Therefore laboratory studies were performed to re-create one of the three processes that were used to separate the plutonium from spent fuel and that created most of the wastes disposed or currently stored in tanks at Hanford. The laboratory simulations were used to compare with current estimates based mainly on flow sheet estimates and spotty historical data. Three simulations of the bismuth phosphate precipitation process show that less that 1% of the Tc-99, Cs-135/137, Sr-90, I-129 carry down with the Pu product and thus these isotopes should have remained within the metals waste streams that after neutralization were sent to single shell tanks. Conversely, these isotopes should not be expected to be found in the first and subsequent cycle waste streams that went to cribs. Measurable quantities (~20 to 30%) of the lanthanides, yttrium, and trivalent actinides (Am and Cm) do precipitate with the Pu product, which is higher than the 10% estimate made for current inventory projections. Surprisingly, Se (added as selenate form) also shows about 10% association with the Pu/bismuth phosphate solids. We speculate that the incorporation of some Se into the bismuth phosphate precipitate is caused by selenate substitution into crystal lattice sites for the phosphate. The bulk of the U daughter product Th-234 and Np-237 daughter product Pa-233 also associate with the solids. We suspect that the Pa daughter products of U (Pa-234 and Pa-231) would also co-precipitate with the bismuth phosphate induced solids. No more than 1 % of the Sr-90 and Sb-125 should carry down with the Pu product that ultimately was purified. Thus the current scheme used to estimate where fission products end up being disposed overestimates by one order of magnitude the partitioning Sr-90, Cs-137, and Sb-125 and by at least two orders of magnitude the portioning of Tc-99 to the first and subsequent cycle waste streams that went to cribs. Conversely, the current scheme underestimates the lanthanide and yttrium fission product quantities that went to cribs by a factor of about 3.

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

    NASA Technical Reports Server (NTRS)

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

    1975-01-01

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

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

    SciTech Connect

    Whitkop, P.G.

    2001-06-26

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

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

    SciTech Connect

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

    2007-04-27

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

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

    DOEpatents

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

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

  5. Nuclear structure and shapes from prompt gamma ray spectroscopy of fission products

    SciTech Connect

    Ahmad, I.; Morss, L.R.; Durell, J.L.

    1996-10-01

    Many nuclear shape phenomena are predicted to occur in neutron-rich nuclei. The best source for the production of these nuclides is the spontaneous fission which produces practically hundreds of nuclides with yields of greater than 0.1 % per decay. Measurements of coincident gamma rays with large Ge arrays have recently been made to obtain information on nuclear structures and shapes of these neutron- rich nuclei. Among the important results that have been obtained from such measurements are octupole correlations in Ba isotopes, triaxial shapes in Ru nuclei, two-phonon vibrations in {sup 106}Mo and level lifetimes and quadrupole moments in Nd isotopes and A=100 nuclei. These data have been used to test theoretical models.

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

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

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

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

    SciTech Connect

    Wichner, R.P. )

    1991-04-01

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

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

    SciTech Connect

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

    2014-10-01

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

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

    SciTech Connect

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

    2013-07-01

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

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

    SciTech Connect

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

    1983-09-01

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

  11. Feasibility of 99Mo production by proton-induced fission of 232Th

    NASA Astrophysics Data System (ADS)

    Abbas, Kamel; Holzwarth, Uwe; Simonelli, Federica; Kozempel, Jan; Cydzik, Izabela; Bulgheroni, Antonio; Cotogno, Giulio; Apostolidis, Christos; Bruchertseifer, Frank; Morgenstern, Alfred

    2012-05-01

    The current global crisis in supply of the medical isotope generator 99Mo/99mTc has triggered much research into alternative non-reactor based production methods for 99Mo including innovative radionuclide production techniques using ion accelerators. A novel method is presented here that has thus far not been considered: 232Th is used as target material to produce carrier-free 99Mo for 99Mo/99mTc generators by proton-induced fission (232Th (p, f) 99Mo). The thick target yields of 99Mo are estimated as 3.6 MBq/?A·h and 21 MBq/?A·h for proton energies of 22 MeV and 40 MeV, respectively, energies that are available from many cyclotrons. With respect to 99Mo reactor based methods using uranium targets, the presented concept using 232Th does not pose proliferation concerns, transport of highly radioactive target materials can be reduced and unused cyclotron capacities could be exploited. Radiochemical target processing could be based on existing technologies of extraction of 99Mo from reactor irradiated 235U. The presented method could be used for co-production of other radioisotopes of medical interest such as 131I.

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

    SciTech Connect

    Gauntt, Randall O.

    2010-04-01

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

  13. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 2013-04-01 false (Item 1204) Oil and gas production, production prices and...S-K Disclosure by Registrants Engaged in Oil and Gas Producing Activities § 229.1204 (Item 1204) Oil and gas production, production prices...

  14. Radioactive Beams from 252Cf Fission Using a Gas Catcher and an ECR Charge Breeder at ATLAS

    SciTech Connect

    Savard, Guy; Pardo, Richard C.; Moore, E. Frank; Hecht, Adam A.; Baker, Sam

    2005-03-15

    A proposed upgrade to the radioactive beam capability of the ATLAS facility has been proposed using 252Cf fission fragments thermalized and collected into a low-energy particle beam using a helium gas catcher. In order to reaccelerate these beams the ATLAS ECR-I will be reconfigured as a charge breeder source. A 1Ci 252Cf source is expected to provide sufficient yield to deliver beams of up to {approx}103 far from stability ions per second on target. A brief facility description and the expected performance information are provided in this report.

  15. Fission Xenon on Mars

    NASA Technical Reports Server (NTRS)

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

    2002-01-01

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

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

    SciTech Connect

    Wright, A.L.

    1994-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

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

  18. Direct irradiation of long-lived fission products in an ATW system

    SciTech Connect

    Carter, Thomas F.; Henderson, Douglass; Sailor, William C.

    1995-09-15

    The feasibility of directly irradiating five long-lived fission products (LLFPs: {sup 79}Se, {sup 93}Zr, {sup 107}Pd, {sup 126}Sn, and {sup 135}Cs, each with a half-life greater than 10,000 years), by incorporating them into the target of an Accelerator Transmutation of Waste (ATW) system is discussed. The important parameters used to judge the feasibility of a direct irradiation system were the target's neutron spallation yield (given in neutrons produced per incident proton), and the removal rate of the LLFP, with the baseline incineration rate set at two light water reactors (LWRs) worth of the LLFP waste per year. A target was constructed which consisted of a LLFP cylindrical ''plug'' inserted into the top (where the proton beam strikes) of a 30 cm radius, 100 cm length lead target. {sup 126}Sn and {sup 79}Se were each found to have high enough removal rates to support two LWR's production of the LLFP per year of ATW operation. For the baseline plug geometry (5 cm radius, 30 cm length) containing {sup 126}Sn, 3.5 LWRs could be supported per year (at 75% beam availability). Furthermore, the addition of a {sup 126}Sn plug had a slightly positive effect on the target's neutron yield. The neutron production was 36.83{+-}.0039 neutrons per proton with a pure lead target having a yield of 36.29{+-}.0038. It was also found that a plug composed of a tin-selenide compound (SnSe) had high enough removal rates to burn two or more reactor years of both LLFPs simultaneously.

  19. Investigation of the recombination losses in a three-electrode cylindrical ionization chamber developed for gamma ray dosimetry of fission product activity

    NASA Astrophysics Data System (ADS)

    Ahmad, N.; Matiullah

    1995-02-01

    A three-electrode ionization chamber has been designed and developed for the gamma ray dosimetry of fission product activity and reported elsewhere. In this paper, the ( I, V) characteristics of the chamber filled, with argon gas at 1.24 MPa (180 psi) pressure, for fission product gamma rays from spent fuel have been studied. To do so, the chamber was irradiated with gamma rays using different numbers of (i.e. up to 4) spent fuel elements. The plateau region is reached above 1200 V and the detector operating voltage is found to be 2 kV. It is observed that in the plateau region the slope increases with an increase in the exposure rate. The ( {1}/{I}, {1}/{V}) and ( I, {1}/{V 2}) characteristic curves reveal the presence of the initial and volume recombination losses. The volume recombination losses are found to be smaller than the initial recombination losses. Both these losses increase with the increasing exposure rate but the increase in the volume recombination losses is slightly greater than that of the initial recombination losses.

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

    SciTech Connect

    Takahashi, Hiroshi; Chen Xinyi; Yu An

    2002-07-01

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

  1. Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses

    SciTech Connect

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

    2014-01-01

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

  2. Flowsheet Testing of the Fission Product Extraction Process as Part of Advanced Aqueous Reprocessing

    SciTech Connect

    Jack Law; Dean R. Peterman; catherine Riddle; David H. Meikrantz; Terry Todd

    2007-06-01

    As part of the Advanced Fuel Cycle Initiative (AFCI), the reduction in volume and heat generation of spent nuclear fuel requiring geologic disposal is currently being addressed. The goal is to optimize utilization of the nation’s first repository and potentially reduce or eliminate the need for additional repositories. This will be achieved through separating long-lived, highly toxic elements, reducing high-level waste volumes and the toxicity of spent nuclear fuel, and reducing the heat generation of spent nuclear fuel. The Idaho National Laboratory (INL) is working closely with a team of national laboratories and other organizations to support development of these separations processes. Key to the reduction of short-term heat load in a geological repository is the separation of 137Cs and 90Sr. Removal of these highly radioactive fission products reduces the short-term (~100 yr) heat generation source of the spent nuclear fuel. Once separated, the Cs and Sr would be placed in storage until the activity has decayed to LLW levels, at which time it could potentially be disposed of as a non-transuranic (TRU) low-level waste (LLW).

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

    SciTech Connect

    Smith, A.B. |; Guenther, P.T.

    1993-09-01

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

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

    SciTech Connect

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

    2012-01-01

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

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

    PubMed

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

    2012-10-01

    We present measurement results of airborne fission products in Chapel Hill, NC, USA, from 62 d following the March 11, 2011, accident at the Fukushima Dai-ichi nuclear power plant. Airborne particle samples were collected daily in air filters and radio-assayed with two high-purity germanium (HPGe) detectors. The fission products (131)I and (137)Cs were measured with maximum activity concentrations of 4.2 ± 0.6 mBq/m(3) and 0.42 ± 0.07 mBq/m(3) respectively. Additional activity from (131,132)I, (134,136,137)Cs and (132)Te were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF). PMID:22348994

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

    E-print Network

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

    2004-01-01

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

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

    SciTech Connect

    Takahashi, H.; Yu, A.

    1997-04-01

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

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

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

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

  9. Fission Product Gamma-Ray Line Pairs Sensitive to Fissile Material and Neutron Energy

    SciTech Connect

    Marrs, R E; Norman, E B; Burke, J T; Macri, R A; Shugart, H A; Browne, E; Smith, A R

    2007-11-15

    The beta-delayed gamma-ray spectra from the fission of {sup 235}U, {sup 238}U, and {sup 239}Pu by thermal and near-14-MeV neutrons have been measured for delay times ranging from 1 minute to 14 hours. Spectra at all delay times contain sets of prominent gamma-ray lines with intensity ratios that identify the fissile material and distinguish between fission induced by low-energy or high-energy neutrons.

  10. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ...to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production...as such, should be included in natural gas sales. Instruction 3 to...

  11. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ...to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production...as such, should be included in natural gas sales. Instruction 3 to...

  12. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ...to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production...as such, should be included in natural gas sales. Instruction 3 to...

  13. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ...to Item 1204: Production of natural gas should include only marketable production of natural gas on an “as sold” basis. Production...as such, should be included in natural gas sales. Instruction 3 to...

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

    NASA Astrophysics Data System (ADS)

    Honaiser, Eduardo Henrique Rangel

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

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

    SciTech Connect

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

    1980-04-01

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

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

    SciTech Connect

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

    2010-01-01

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

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

    SciTech Connect

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

    2000-05-11

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

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

    SciTech Connect

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

    2010-11-01

    All mass streams from separations and fuel fabrication are products that must meet criteria. Those headed for disposal must meet waste acceptance criteria (WAC) for the eventual disposal sites corresponding to their waste classification. Those headed for reuse must meet fuel or target impurity limits. A “loss” is any material that ends up where it is undesired. The various types of losses are linked in the sense that as the loss of transuranic (TRU) material into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. We have analyzed four separation options and two fuel fabrication options in a generic fuel cycle. The separation options are aqueous uranium extraction plus (UREX+1), electrochemical, Atomics International reduction oxidation separation (AIROX), and melt refining. UREX+1 and electrochemical are traditional, full separation techniques. AIROX and melt refining are taken as examples of limited separations, also known as minimum fuel treatment. The fuels are oxide and metal. To define a generic fuel cycle, a fuel recycling loop is fed from used light water reactor (LWR) uranium oxide fuel (UOX) at 51 MWth-day/kg-iHM burnup. The recycling loop uses a fast reactor with TRU conversion ratio (CR) of 0.50. Excess recovered uranium is put into storage. Only waste, not used fuel, is disposed – unless the impurities accumulate to a level so that it is impossible to make new fuel for the fast reactor. Impurities accumulate as dictated by separation removal and fission product generation. Our model approximates adjustment to fast reactor fuel stream blending of TRU and U products from incoming LWR UOX and recycling FR fuel to compensate for impurity accumulation by adjusting TRU:U ratios. Our mass flow model ignores postulated fuel impurity limits; we compare the calculated impurity values with those limits to identify elements of concern. AIROX and melt refining cannot be used to separate used LWR UOX-51 because they cannot separate U from TRU, it is then impossible to make X% TRU for fast reactors with UOX-51 used fuel with 1.3% TRU. AIROX and melt refining can serve in the recycle loop for about 3 recycles, at which point the accumulated impurities displace fertile uranium and the fuel can no longer be as critical as the original fast reactor fuel recipe. UREX+1 and electrochemical can serve in either capacity; key impurities appear to be lanthanides and several transition metals.

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

    SciTech Connect

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

    2009-02-01

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

  1. Shale gas production: potential versus actual greenhouse gas emissions

    NASA Astrophysics Data System (ADS)

    O'Sullivan, Francis; Paltsev, Sergey

    2012-12-01

    Estimates of greenhouse gas (GHG) emissions from shale gas production and use are controversial. Here we assess the level of GHG emissions from shale gas well hydraulic fracturing operations in the United States during 2010. Data from each of the approximately 4000 horizontal shale gas wells brought online that year are used to show that about 900 Gg CH4 of potential fugitive emissions were generated by these operations, or 228 Mg CH4 per well—a figure inappropriately used in analyses of the GHG impact of shale gas. In fact, along with simply venting gas produced during the completion of shale gas wells, two additional techniques are widely used to handle these potential emissions: gas flaring and reduced emission ‘green’ completions. The use of flaring and reduced emission completions reduce the levels of actual fugitive emissions from shale well completion operations to about 216 Gg CH4, or 50 Mg CH4 per well, a release substantially lower than several widely quoted estimates. Although fugitive emissions from the overall natural gas sector are a proper concern, it is incorrect to suggest that shale gas-related hydraulic fracturing has substantially altered the overall GHG intensity of natural gas production.

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

    NASA Astrophysics Data System (ADS)

    Zhu, Sha

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

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

    NASA Astrophysics Data System (ADS)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

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

    E-print Network

    Grover, Tarun

    2008-10-10

    by the formation of secondary hydrates during gas production. I used the coupled fluid flow and geomechanical model “TOUGH+Hydrate- FLAC3D” to model geomechanical performance during gas production from hydrates in an offshore hydrate deposit. I modeled...+H)............................................................. 73 4.3 TOUGH+Hydrate-FLAC3D (T+F)............................................. 81 V RESERVOIR PERFORMANCE OF THE MESSOYAKHA FIELD . 86 5.1 Introduction ................................................................................. 86...

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

    SciTech Connect

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

    2014-11-01

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

  6. FISSION AND FUSION Lecture 11.VII

    E-print Network

    Smith, Nathanael J.

    FISSION AND FUSION Lecture 11.VII #12;Review 2 Why does a system of nucleons form a nucleus, rather nucleon Overview 4 Tipler #12;Nuclear Fission Nuclear fission is a special type of nuclear reaction in which a heavy nucleus captures a neutron, and splits ("fissions") into two, roughly equal mass, products

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

    SciTech Connect

    Gohar, Y.

    2000-11-15

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

  8. How EIA Estimates Natural Gas Production

    EIA Publications

    2004-01-01

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

  9. The preparation and characterization of a Ni-3 wt.% U reference fission product source for super phenix

    NASA Astrophysics Data System (ADS)

    Crouzet, J.; De Wilde, L.; Pauwels, J.; Van Audenhove, J.

    1987-06-01

    The fabrication and quality control of Ni-3 wt.% U (93% 235U) alloy rods (diameter: 8 mm, total length: 8 m) used as a reference fission product source (RFPS) are described. This RFPs is needed for the calibration of the fuel element rupture detector which is used for the continuous monitoring of the delayed neutron activity in the primary sodium of the Super Phenix fast neutron reactor. Special care was taken to obtain a sufficient alloy purity, a bright surface finish and a homogeneous 235U content on the surface of the rods. Finally the usefulness of the RFPS in Super Phenix has been demonstrated.

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

    SciTech Connect

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

    2006-07-01

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

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

    SciTech Connect

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

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste stream options in terms of waste loading and/or decay time required before treatment. For Option 1, glass ceramics show an increase in waste loading of 15 mass % and reduction in decay time of 24 years. Decay times of {approx}50 years or longer are close to the expected age of the fuel that will be reprocessed when the modified open or closed fuel cycle is expected to be put into action. Option 2 shows a 2x to 2.5x increase in waste loading with decay times of only 45 years. Note that for Option 2 glass, the required decay time before treatment is only 35 years because of the waste loading limits related to the solubility of MoO{sub 3} in glass. If glass was evaluated for similar waste loadings as those achieved in Option 2 glass ceramics, the decay time would be significantly longer than 45 years. These glass ceramics are not optimized, but already they show the potential to dramatically reduce the amount of waste generated while still utilizing the proven processing technology used for glass production.

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

    PubMed

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

    2015-11-17

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

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

    SciTech Connect

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

    1987-01-01

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

  14. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

    PubMed

    Koš?ál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojt?ch; Mil?ák, Ján

    2015-11-01

    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins. PMID:26351013

  15. ConocoPhillips Gas Hydrate Production Test

    SciTech Connect

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

    2013-06-30

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

  16. Powering the World: Offshore Oil & Gas Production

    E-print Network

    Patzek, Tadeusz W.

    Powering the World: Offshore Oil & Gas Production Macondo post-blowout operations Tad Patzek of a banking account) is mistakenly equated with the speed of drawing it down (ATM withdrawals) Few understand

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

    SciTech Connect

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

    2011-09-23

    Glass-ceramic waste form development began in FY 2010 examining two combined waste stream options: (1) alkaline earth (CS) + lanthanide (Ln), and (2) + transition metal (TM) fission-product waste streams generated by the uranium extraction (UREX+) separations process. Glass-ceramics were successfully developed for both options however; Option 2 was selected over Option 1, at the conclusion of 2010, because Option 2 immobilized all three waste streams with only a minimal decrease in waste loading. During the first year, a series of three glass (Option 2) were fabricated that varied waste loading-WL (42, 45, and 50 mass%) at fixed molar ratios of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali both at 1.75. These glass-ceramics were slow cooled and characterized in terms of phase assemblage and preliminary irradiation stability. This fiscal year, further characterization was performed on the FY 2010 Option 2 glass-ceramics in terms of: static leach testing, phase analysis by transmission electron microscopy (TEM), and irradiation stability (electron and ion). Also, a new series of glass-ceramics were developed for Option 2 that varied the additives: Al{sub 2}O{sub 3} (0-6 mass%), molar ratio of CaO/MoO{sub 3} and B{sub 2}O{sub 3}/alkali (1.75 to 2.25) and waste loading (50, 55, and 60 mass%). Lastly, phase pure powellite and oxyapatite were synthesized for irradiation studies. Results of this fiscal year studies showed compositional flexibility, chemical stability, and radiation stability in the current glass-ceramic system. First, the phase assemblages and microstructure of all of the FY 2010 and 2011 glass-ceramics are very similar once subjected to the slow cool heat treatment. The phases identified in these glass-ceramics were oxyapatite, powellite, cerianite, and ln-borosilicate. This shows that variations in waste loading or additives can be accommodated without drastically changing the phase assemblage of the waste form, thus making the processing and performance characteristics of the waste form more predictable/flexible. However, in the future, the glass phase still needs to be accurately characterized to determine the effects of waste loading and additives on the glass structure. Initial investigations show a borosilicate glass phase rich in silica. Second, the normalized concentrations of elements leached from the waste form during static leach testing were all below 0.6 g/L after 28d at 90 C, by the Product Consistency Test (PCT), method B. These normalized concentrations are on par with durable waste glasses such as the Low-Activity Reference Material (LRM) glass. The release rates for the crystalline phases (oxyapatite and powellite) appear to be lower (more durable) than the glass phase based on the relatively low release rates of Mo, Ca, and Ln found in the crystalline phases compared to Na and B that are mainly observed in the glass phase. However, further static leach testing on individual crystalline phases is needed to confirm this statement. Third, Ion irradiation and In situ TEM observations suggest that these crystalline phases (such as oxyapatite, ln-borosilicate, and powellite) in silicate based glass ceramic waste forms exhibit stability to 1000 years at anticipated doses (2 x 10{sup 10}-2 x 10{sup 11} Gy). This is adequate for the short lived isotopes in the waste, which lead to a maximum cumulative dose of {approx}7 x 10{sup 9} Gy, reached after {approx}100 yrs, beyond which the dose contributions are negligible. The cumulate dose calculations are based on a glass-ceramic at WL = 50 mass%, where the fuel has a burn-up of 51GWd/MTIHM, immobilized after 5 yr decay from reactor discharge.

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

    SciTech Connect

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

    2011-01-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He; Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl; K; Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [H.

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

    NASA Astrophysics Data System (ADS)

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

    2011-12-01

    The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, "ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology," Nuclear Data Sheets 107, 2931 (2006)].

  20. Benchmarking the LAHET fission models

    SciTech Connect

    Prael, R.E.

    1995-12-31

    There has been considerable interest in improving the fission models in the LAHET Monte Carlo code for the transport and interaction of nucleons, pions, muons, fight ions, and antinucleons. Although subactinide fission contributes little to neutron production in lead or tungsten targets, it can be significant for simulation of target activation and fission product contamination. The availability of new data permits new comparisons to be made between experiment and calculation.

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

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  4. Exact Solution of Fractional Diffusion Model with Source Term used in Study of Concentration of Fission Product in Uranium Dioxide Particle

    NASA Astrophysics Data System (ADS)

    Fang, Chao; Cao, Jian-Zhu; Sun, Li-Feng

    2011-05-01

    The exact solution of fractional diffusion model with a location-independent source term used in the study of the concentration of fission product in spherical uranium dioxide (UO2) particle is built. The adsorption effect of the fission product on the surface of the UO2 particle and the delayed decay effect are also considered. The solution is given in terms of Mittag—Leffler function with finite Hankel integral transformation and Laplace transformation. At last, the reduced forms of the solution under some special physical conditions, which is used in nuclear engineering, are obtained and corresponding remarks are given to provide significant exact results to the concentration analysis of nuclear fission products in nuclear reactor.

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

    SciTech Connect

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

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the international nuclear data community as of March 2005. The accuracy of the cross-section data was investigated by comparing existing cross-section evaluations against available measured cross-section data. When possible, benchmark calculations were also used to assess the performance of the latest FP cross-section data. Since March 2005, the U.S. and European data projects have released newer versions of their respective data files. Although there have been updates to the international data files and to some degree FP data, much of the updates have included nuclear cross-section modeling improvements at energies above the resonance region. The one exception is improved ENDF/B-VII cross-section uncertainty data or covariance data for gadolinium isotopes. In particular, ENDF/B-VII includes improved 155Gd resonance parameter covariance data, but they are based on previously measured resonance data. Although the new covariance data are available for 155Gd, the conclusions of the FP cross-section data assessment of this report still hold in lieu of the newer international cross-section data files. Based on the FP data assessment, there is judged to be a need for new total and capture cross-section measurements and corresponding cross-section evaluations, in a prioritized manner, for the nine FPs to provide the improved information and technical rigor needed for criticality safety analyses.

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

    SciTech Connect

    Clark, E.

    2013-08-31

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

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

    E-print Network

    M. Veselsky

    2003-02-11

    The mechanism of fusion hindrance, an effect observed in the reactions of cold, warm and hot fusion leading to production of the superheavy elements, is investigated. A systematics of transfermium production cross sections is used to determine fusion probabilities. Mechanism of fusion hindrance is described as a competition of fusion and quasi-fission. Available evaporation residue cross sections in the superheavy region are reproduced satisfactorily. Analysis of the measured capture cross sections is performed and a sudden disappearance of the capture cross sections is observed at low fusion probabilities. A dependence of the fusion hindrance on the asymmetry of the projectile-target system is investigated using the available data. The most promising pathways for further experiments are suggested.

  8. Isomers in Fission Fragments

    SciTech Connect

    Urban, W.; Faust, H.; Jentschel, M.; Koester, U.; Krempel, J.; Materna, Th.; Mutti, P.; Soldner, T.; Genevey, J.; Pinston, J. A.; Simpson, G.; Sieja, K.; Nowacki, F.; Dorvaux, O.; Gall, B. J. P.; Roux, B.; Dare, J. A.

    2009-01-28

    The structure of neutron-rich nuclei produced as secondary fission fragments was investigated using the EUROGAM and GAMMASPHERE ACS arrays, the LOHENGRIN fission-fragment mass separator and the FIFI fission-fragment identifier. Fission products were populated in spontaneous fission of {sup 248}Cm and {sup 252}Cf and in thermal neutron-induced fission of {sup 233}U, {sup 235}U and {sup 241}Pu at ILL Grenoble. Particularly useful in such studies are isomeric states, well populated in fission due to their yrast character, easy to detect due to their long half lives and easy to interpret because of their relatively simple composition. We discuss their role in studies of neutron-rich nuclei, giving examples of isomers found in our recent experiments. A special type of K-isomers, resulting from 'crossing' of extruder and intruder orbitals plays a role in the mechanism of a sudden onset of deformation in the A = 100 and A = 150 regions. We present evidence for these isomers in both regions. Possible further studies in this field are proposed.

  9. Myanmar production meets first-gas targets

    SciTech Connect

    Lepage, A.

    1998-09-07

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

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

    SciTech Connect

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-30

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo{sup 99} used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 10{sup 6} cm{sup ?1}) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 10{sup 6} cm{sup ?1}) the limits were exceeded.

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

    NASA Astrophysics Data System (ADS)

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

    2014-08-01

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

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

    SciTech Connect

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

    2014-03-01

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

  13. Production Pathways and Separation Procedures for High-Diagnostic-Value Activation Species, Fission Products, and Actinides Required for Preparation of Realistic Synthetic Post-Detonation Nuclear Debris

    SciTech Connect

    Faye, S A; Shaughnessy, D A

    2015-08-19

    The objective of this project is to provide a comprehensive study on the production routes and chemical separation requirements for activation products, fission products, and actinides required for the creation of realistic post-detonation surrogate debris. Isotopes that have been prioritized by debris diagnosticians will be examined for their ability to be produced at existing irradiation sources, production rates, and availability of target materials, and chemical separation procedures required to rapidly remove the products from the bulk target matrix for subsequent addition into synthetic debris samples. The characteristics and implications of the irradiation facilities on the isotopes of interest will be addressed in addition to a summary of the isotopes that are already regularly produced.

  14. Mitochondrial fusion but not fission regulates larval growth and synaptic development through steroid hormone production

    PubMed Central

    Sandoval, Hector; Yao, Chi-Kuang; Chen, Kuchuan; Jaiswal, Manish; Donti, Taraka; Lin, Yong Qi; Bayat, Vafa; Xiong, Bo; Zhang, Ke; David, Gabriela; Charng, Wu-Lin; Yamamoto, Shinya; Duraine, Lita; Graham, Brett H; Bellen, Hugo J

    2014-01-01

    Mitochondrial fusion and fission affect the distribution and quality control of mitochondria. We show that Marf (Mitochondrial associated regulatory factor), is required for mitochondrial fusion and transport in long axons. Moreover, loss of Marf leads to a severe depletion of mitochondria in neuromuscular junctions (NMJs). Marf mutants also fail to maintain proper synaptic transmission at NMJs upon repetitive stimulation, similar to Drp1 fission mutants. However, unlike Drp1, loss of Marf leads to NMJ morphology defects and extended larval lifespan. Marf is required to form contacts between the endoplasmic reticulum and/or lipid droplets (LDs) and for proper storage of cholesterol and ecdysone synthesis in ring glands. Interestingly, human Mitofusin-2 rescues the loss of LD but both Mitofusin-1 and Mitofusin-2 are required for steroid-hormone synthesis. Our data show that Marf and Mitofusins share an evolutionarily conserved role in mitochondrial transport, cholesterol ester storage and steroid-hormone synthesis. DOI: http://dx.doi.org/10.7554/eLife.03558.001 PMID:25313867

  15. Observation of new microsecond isomers among fission products of 345 MeV/nucleon 238U

    E-print Network

    D. Kameda; T. Kubo; T. Ohnishi; K. Kusaka; A. Yoshida; K. Yoshida; M. Ohtake; N. Fukuda; H. Takeda; K. Tanaka; N. Inabe; Y. Yanagisawa; Y. Gono; H. Watanabe; H. Otsu; H. Baba; T. Ichihara; Y. Yamaguchi; M. Takechi; S. Nishimura; H. Ueno; A. Yoshimi; H. Sakurai; T. Motobayashi; T. Nakao; Y. Mizoi; M. Matsushita; K. Ieki; N. Kobayashi; K. Tanaka; Y. Kawada; N. Tanaka; S. Deguchi; Y. Satou; Y. Kondo; T. Nakamura; K. Yoshinaga; C. Ishii; H. Yoshii; Y. Miyashita; N. Uematsu; Y. Shiraki; T. Sumikama; J. Chiba; E. Ideguchi; A. Saito; T. Yamaguchi; I. Hachiuma; T. Suzuki; T. Moriguchi; A. Ozawa; T. Ohtsubo; M. A. Famiano; H. Geissel; A. S. Nettleton; O. B. Tarasov; D. Bazin; B. M. Sherrill; S. L. Manikonda; J. A. Nolen

    2012-11-08

    A search for isomeric gamma-decays among fission fragments from 345 MeV/nucleon 238U has been performed at the RIKEN Nishina Center RI Beam Factory. Fission fragments were selected and identified using the superconducting in-flight separator BigRIPS and were implanted in an aluminum stopper. Delayed gamma-rays were detected using three clover-type high-purity germanium detectors located at the focal plane within a time window of 20 microseconds following the implantation. We identified a total of 54 microsecond isomers with half-lives of ~0.1 - 10 microseconds, including discovery of 18 new isomers in very neutron-rich nuclei: 59Tim, 90Asm, 92Sem, 93Sem, 94Brm, 95Brm, 96Brm, 97Rbm, 108Nbm, 109Mom, 117Rum, 119Rum, 120Rhm, 122Rhm, 121Pdm, 124Pdm, 124Agm and 126Agm, and obtained a wealth of spectroscopic information such as half-lives, gamma-ray energies, gamma-ray relative intensities and gamma-gamma coincidences over a wide range of neutron-rich exotic nuclei. Proposed level schemes are presented for 59Tim, 82Gam, 92Brm, 94Brm, 95Brm, 97Rbm, 98Rbm, 108Nbm, 108Zrm, 109Mom, 117Rum, 119Rum, 120Rhm, 122Rhm, 121Pdm, 124Agm and 125Agm, based on the obtained spectroscopic information and the systematics in neighboring nuclei. Nature of the nuclear isomerism is discussed in relation to evolution of nuclear structure.

  16. Spectroscopy of few-particle nuclei around magic {sup 132}Sn from fission product {gamma}-ray studies.

    SciTech Connect

    Zhang, C. T.

    1998-07-29

    We are studying the yrast structure of very neutron-rich nuclei around doubly magic {sup 132}Sn by analyzing fission product {gamma}-ray data from a {sup 248}Cm source at Eurogam II. Yrast cascades in several few-valence-particle nuclei have been identified through {gamma}{gamma} cross coincidences with their complementary fission partners. Results for two-valence-particle nuclei {sup 132}Sb, {sup 134}Te, {sup 134}Sb and {sup 134}Sn provide empirical nucleon-nucleon interactions which, combined with single-particle energies already known in the one-particle nuclei, are essential for shell-model analysis in this region. Findings for the N = 82 nuclei {sup 134}Te and {sup 135}I have now been extended to the four-proton nucleus {sup 136}Xe. Results for the two-neutron nucleus {sup 134}Sn and the N = 83 isotones {sup 134}Sb, {sup 135}Te and {sup 135}I open up the spectroscopy of nuclei in the northeast quadrant above {sup 132}Sn.

  17. JASPER [Japanese-American Shielding Program of Experimental Research], USDOE/PNC shielding research program: Analysis of the JASPER fission gas plenum experiment

    SciTech Connect

    Slater, C.O.

    1990-05-01

    The results of the analysis of the Fission Gas Plenum Experiment are presented. This experiment is the second in a series of several experiments comprising a joint US DOE-Japan PNC Shielding Research Program (JASPER). The four Fission Gas Plenum Experiment configurations, designed for the measurement of neutron streaming through the fission gas plenum region, were analyzed using Monte Carlo and two-dimensional discrete ordinated methods. Calculated results compared well with measured results in many cases, although results were consistently underpredicted for the shorter plenum configurations. Like the measured data, the calculated results indicated no significant streaming when results from the heterogeneous mockups were compared to those from the homogeneous mockups. An explanation is given as to why little streaming was observed. The Hornyak button dose rates were overpredicted because of a normalization problem with the response function but yielded horizontal traverse curves whose shapes agreed well with the measured shapes to the same extent as did those for the other integral detectors. 16 refs., 16 figs., 4 tabs.

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

    SciTech Connect

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

    2011-01-01

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.

  19. Yrast excitations around doubly magic {sup 132}Sn from fission product {gamma}-ray studies.

    SciTech Connect

    Zhang, C. T.; Bhattacharyya, P.; Daly, P. J.; Broda, R.; Grabowski, Z. W.; Nisius, D.; Ahmad, I.; Ishii, T.; Carpenter, M. P.; Morss, L. R.; Phillips, W. R.; Durell, J. L.; Leddy, M. J.; Smith, A. G.; Urban, W.; Varley, B. J.; Schulz, N.; Lubkiewicz, E.; Bentaleb, M.; Blomqvist, J.; Physics; Purdue Univ.; Univ. of Manchester; Univ. of Louis Pasteur; Royal Inst. of Tech.

    1996-10-01

    Prompt {gamma}-ray cascades in neutron-rich nuclei around doubly magic {sup 132}Sn have been studied at Eurogam II using a {sup 248}Cm fission source. Yrast states to above 5.5 MeV in the two- and three-proton N = 82 isotones {sup 134}Te and {sup 135}I are reported. They are interpreted in terms of valence proton and particle-hole core excitations with the help of shell model calculations employing empirical nucleon-nucleon interactions from both {sup 132}Sn and {sup 208}Pb regions. A serious inconsistency in the accepted masses of N = 82 isotones near {sup 132}Sn is discovered but not resolved.

  20. Production of CoQ10 in fission yeast by expression of genes responsible for CoQ10 biosynthesis.

    PubMed

    Moriyama, Daisuke; Hosono, Kouji; Fujii, Makoto; Washida, Motohisa; Nanba, Hirokazu; Kaino, Tomohiro; Kawamukai, Makoto

    2015-01-01

    Coenzyme Q10 (CoQ10) is essential for energy production and has become a popular supplement in recent years. In this study, CoQ10 productivity was improved in the fission yeast Schizosaccharomyces pombe. Ten CoQ biosynthetic genes were cloned and overexpressed in S. pombe. Strains expressing individual CoQ biosynthetic genes did not produce higher than a 10% increase in CoQ10 production. In addition, simultaneous expression of all ten coq genes did not result in yield improvements. Genes responsible for the biosynthesis of p-hydroxybenzoate and decaprenyl diphosphate, both of which are CoQ biosynthesis precursors, were also overexpressed. CoQ10 production was increased by overexpression of Eco_ubiC (encoding chorismate lyase), Eco_aroF(FBR) (encoding 3-deoxy-D-arabino-heptulosonate 7-phosphate synthase), or Sce_thmgr1 (encoding truncated HMG-CoA reductase). Furthermore, simultaneous expression of these precursor genes resulted in two fold increases in CoQ10 production. PMID:25647499

  1. Deep Atomic Binding (DAB) Hypothesis: A New Approach of Fission Product Chemistry

    SciTech Connect

    Ajlouni, Abdul-Wali M.S.

    2006-07-01

    Former studies assumed that, after fission process occurs, the highly ionized new born atoms (20-22 positive charge), ionize the media in which they pass through before becoming stable atoms in a manner similar to 4-MeV ?-particles. Via ordinary chemical reactions with the surroundings, each stable atom has a probability to form chemical compound. Since there are about 35 different elemental atoms created through fission processes, a large number of chemical species were suggested to be formed. But, these suggested chemical species were not found in the environment after actual releases of FP during accidents like TMI (USA, 1979), and Chernobyl (former USSR, 1986), also the models based on these suggested reactions and species could not interpret the behavior of these actual species. It is assumed here that the ionization states of the new born atoms and the long term high temperature were not dealt with in an appropriate way and they were the reasons of former models failure. Our new approach of Deep Atomic Binding (DAB) based on the following: 1-The new born atoms which are highly ionized, 10-12 electrons associated with each nucleus, having a large probability to create bonds between them to form molecules. These bonds are at the L, or M shells, and we call it DAB. 2-The molecules stay in the reactor at high temperatures for long periods, so they undergo many stages of composition and decomposition to form giant molecules. By applying DAB approach, field data from Chernobyl, TMI and nuclear detonations could be interpreted with a wide coincidence resulted. (author)

  2. Collection of fission and activation product elements from fresh and ocean waters: a comparison of traditional and novel sorbents

    SciTech Connect

    Johnson, Bryce E.; Santschi, Peter H.; Addleman, Raymond S.; Douglas, Matthew; Davidson, Joseph D.; Fryxell, Glen E.; Schwantes, Jon M.

    2010-04-01

    Monitoring natural waters for the inadvertent release of radioactive fission products produced as a result of nuclear power generation downstream from these facilities is essential for maintaining water quality. To this end, we evaluated sorbents for simultaneous in-situ large volume extraction of radionuclides with both soft (e.g., Ag) and hard metal (e.g., Co, Zr, Nb, Ba, and Cs) or anionic (e.g., Ru, Te, Sb) character. In this study, we evaluated a number of conventional and novel nanoporous sorbents in both fresh and salt waters. In most cases, the nanoporous sorbents demonstrated enhanced retention of analytes. Salinity had significant effects upon sorbent performance and was most significant for hard cations, specifically Cs and Ba. The presence of natural organic matter had little effect on the ability of chemisorbents to extract target elements.

  3. Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products.

    PubMed

    Koš?ál, Michal; Švadlenková, Marie; Mil?ák, Ján; Rypar, Vojt?ch; Koleška, Michal

    2014-07-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well. PMID:24566373

  4. Collection of fission and activation product elements from fresh and ocean waters: a comparison of traditional and novel sorbents.

    PubMed

    Johnson, Bryce E; Santschi, Peter H; Addleman, Raymond Shane; Douglas, Matt; Davidson, Joseph D; Fryxell, Glen E; Schwantes, Jon M

    2011-01-01

    Monitoring natural waters for the inadvertent release of radioactive fission products produced as a result of nuclear power generation downstream from these facilities is essential for maintaining water quality. To this end, we evaluated sorbents for simultaneous in-situ large volume extraction of radionuclides with both soft (e.g., Ag) and hard metal (e.g., Co, Zr, Nb, Ba, and Cs) or anionic (e.g., Ru, Te, Sb) character. In this study, we evaluated a number of conventional and novel nanoporous sorbents in both fresh and salt waters. In most cases, the nanoporous sorbents demonstrated enhanced retention of analytes. Salinity had significant effects upon sorbent performance and was most significant for hard cations, specifically Cs and Ba. The presence of natural organic matter had little effect on the ability of chemisorbents to extract target elements. PMID:20870414

  5. Thermal transport in UO2 with defects and fission products by molecular dynamics simulations

    SciTech Connect

    Liu, Xiang-Yang; Cooper, Michael William Donald; Mcclellan, Kenneth James; Lashley, Jason Charles; Byler, Darrin David; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-14

    The importance of the thermal transport in nuclear fuel has motivated a wide range of experimental and modelling studies. In this report, the reduction of thermal transport in UO2 due to defects and fission products has been investigated using non-equilibrium MD simulations, with two sets of empirical potentials for studying the degregation of UO2 thermal conductivity including a Buckingham type interatomic potential and a recently developed EAM type interatomic potential. Additional parameters for U5+ and Zr4+ in UO2 have been developed for the EAM potential. The thermal conductivity results from MD simulations are then corrected for the spin-phonon scattering through Callaway model formulations. To validate the modelling results, comparison was made with experimental measurements on single crystal hyper-stoichiometric UO2+x samples.

  6. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    SciTech Connect

    Henager, Charles H.; Jiang, Weilin

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

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

    E-print Network

    U. Forsberg; D. Rudolph; L. -L. Andersson; A. Di Nitto; Ch. E. Düllmann; J. M. Gates; P. Golubev; K. E. Gregorich; C. J. Gross; R. -D. Herzberg; F. P. Hessberger; J. Khuyagbaatar; J. V. Kratz; K. Rykaczewski; L. G. Sarmiento; M. Schädel; A. Yakushev; S. Åberg; D. Ackermann; M. Block; H. Brand; B. G. Carlsson; D. Cox; X. Derkx; J. Dobaczewski; K. Eberhardt; J. Even; C. Fahlander; J. Gerl; E. Jäger; B. Kindler; J. Krier; I. Kojouharov; N. Kurz; B. Lommel; A. Mistry; C. Mokry; W. Nazarewicz; H. Nitsche; J. P. Omtvedt; P. Papadakis; I. Ragnarsson; J. Runke; H. Schaffner; B. Schausten; Y. Shi; P. Thörle-Pospiech; T. Torres; T. Traut; N. Trautmann; A. Türler; A. Ward; D. E. Ward; N. Wiehl

    2015-02-10

    Products of the fusion-evaporation reaction Ca-48 + Am-243 were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum f\\"ur Schwerionenforschung. Amongst the detected thirty correlated alpha-decay chains associated with the production of element Z=115, two recoil-alpha-fission and five recoil-alpha-alpha-fission events were observed. The latter are similar to four such events reported from experiments performed at the Dubna gas-filled separator. Contrary to their interpretation, we propose an alternative view, namely to assign eight of these eleven decay chains of recoil-alpha(-alpha)-fission type to start from the 3n-evaporation channel 115-288. The other three decay chains remain viable candidates for the 2n-evaporation channel 115-289.

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

    E-print Network

    Forsberg, U; Andersson, L -L; Di Nitto, A; Düllmann, Ch E; Gates, J M; Golubev, P; Gregorich, K E; Gross, C J; Herzberg, R -D; Hessberger, F P; Khuyagbaatar, J; Kratz, J V; Rykaczewski, K; Sarmiento, L G; Schädel, M; Yakushev, A; Åberg, S; Ackermann, D; Block, M; Brand, H; Carlsson, B G; Cox, D; Derkx, X; Dobaczewski, J; Eberhardt, K; Even, J; Fahlander, C; Gerl, J; Jäger, E; Kindler, B; Krier, J; Kojouharov, I; Kurz, N; Lommel, B; Mistry, A; Mokry, C; Nazarewicz, W; Nitsche, H; Omtvedt, J P; Papadakis, P; Ragnarsson, I; Runke, J; Schaffner, H; Schausten, B; Shi, Y; Thörle-Pospiech, P; Torres, T; Traut, T; Trautmann, N; Türler, A; Ward, A; Ward, D E; Wiehl, N

    2015-01-01

    Products of the fusion-evaporation reaction Ca-48 + Am-243 were studied with the TASISpec set-up at the gas-filled separator TASCA at the GSI Helmholtzzentrum f\\"ur Schwerionenforschung. Amongst the detected thirty correlated alpha-decay chains associated with the production of element Z=115, two recoil-alpha-fission and five recoil-alpha-alpha-fission events were observed. The latter are similar to four such events reported from experiments performed at the Dubna gas-filled separator. Contrary to their interpretation, we propose an alternative view, namely to assign eight of these eleven decay chains of recoil-alpha(-alpha)-fission type to start from the 3n-evaporation channel 115-288. The other three decay chains remain viable candidates for the 2n-evaporation channel 115-289.

  9. I-NERI ANNUAL TECHNICAL PROGRESS REPORT: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels

    SciTech Connect

    S. Frank

    2009-09-01

    An attractive alternative to the once-through disposal of electrorefiner salt is to selectively remove the active fission products from the salt and recycle the salt back to the electrorefiner (ER). This would allow salt reuse for some number of cycles before ultimate disposal of the salt in a ceramic waste form. Reuse of ER salt would, thus, greatly reduce the volume of ceramic waste produced during the pyroprocessing of spent nuclear fuel. This final portion of the joint I-NERI research project is to demonstrate the separation of fission products from molten ER salt by two methods previously selected during phase two (FY-08) of this project. The two methods selected were salt/zeolite contacting and rare-earth fission product precipitation by oxygen bubbling. The ER salt used in these tests came from the Mark-IV electrorefiner used to anodically dissolved driver fuel from the EBR-II reactor on the INL site. The tests were performed using the Hot Fuel Dissolution Apparatus (HFDA) located in the main cell of the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels complex on the INL site. Results from these tests were evaluated during a joint meeting of KAERI and INL investigators to provide recommendations as to the future direction of fission product removal from electrorefiner salt that accumulate during spent fuel treatment. Additionally, work continued on kinetic measurements of surrogate quaternary salt systems to provide fundamental kinetics on the ion exchange system and to expand the equilibrium model system developed during the first two phases of this project. The specific objectives of the FY09 I-NERI research activities at the INL include the following: • Perform demonstration tests of the selected KAERI precipitation and INL salt/zeolite contacting processes for fission product removal using radioactive, fission product loaded ER salt • Continue kinetic studies of the quaternary Cs/Sr-LiCl-KCl system to determine the rate of ion exchange during the salt/zeolite contacting process • Compare the adsorption models to experimentally obtained, ER salt results • Evaluate results obtained from the oxygen precipitation and salt/zeolite ion exchange studies to determine the best processes for selective fission-product removal from electrorefiner salt.

  10. Feasibility of an on-line fission-gas-leak detection system

    NASA Technical Reports Server (NTRS)

    Lustig, P. H.

    1973-01-01

    Calculations were made to determine if a cladding failure could be detected in a 100-kW zirconium hydride reactor primary system by monitoring the highly radioactive NaK coolant for the presence of I-131. The system is to be completely sealed. A leak of 0.01 percent from a single fuel pin was postulated. The 0.364-MeV gamma of I-131 could be monitored on an almost continuous basis, while its presence could be varified by using a longer counting time for the 0.638-MeV gamma. A lithium-drifted germanium detector would eliminate radioactive corrosion product interference that could occur with a sodium iodide scintillation detector.

  11. Bio-gas production from alligator weeds

    NASA Technical Reports Server (NTRS)

    Latif, A.

    1976-01-01

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

  12. New Methodology for Natural Gas Production Estimates

    EIA Publications

    2010-01-01

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

  13. Metal powder production by gas atomization

    NASA Technical Reports Server (NTRS)

    Ting, E. Y.; Grant, N. J.

    1986-01-01

    The confined liquid, gas-atomization process was investigated. Results from a two-dimensional water model showed the importance of atomization pressure, as well as delivery tube and atomizer design. The atomization process at the tip of the delivery tube was photographed. Results from the atomization of a modified 7075 aluminum alloy yielded up to 60 wt pct. powders that were finer than 45 microns in diameter. Two different atomizer designs were evaluated. The amount of fine powders produced was correlated to a calculated gas-power term. An optimal gas-power value existed for maximized fine powder production. Atomization at gas-power greater than or less than this optimal value produced coarser powders.

  14. Observation of new microsecond isomers among fission products from in-flight fission of 345 MeV/nucleon 238U

    NASA Astrophysics Data System (ADS)

    Kameda, D.; Kubo, T.; Ohnishi, T.; Kusaka, K.; Yoshida, A.; Yoshida, K.; Ohtake, M.; Fukuda, N.; Takeda, H.; Tanaka, K.; Inabe, N.; Yanagisawa, Y.; Gono, Y.; Watanabe, H.; Otsu, H.; Baba, H.; Ichihara, T.; Yamaguchi, Y.; Takechi, M.; Nishimura, S.; Ueno, H.; Yoshimi, A.; Sakurai, H.; Motobayashi, T.; Nakao, T.; Mizoi, Y.; Matsushita, M.; Ieki, K.; Kobayashi, N.; Tanaka, K.; Kawada, Y.; Tanaka, N.; Deguchi, S.; Satou, Y.; Kondo, Y.; Nakamura, T.; Yoshinaga, K.; Ishii, C.; Yoshii, H.; Miyashita, Y.; Uematsu, N.; Shiraki, Y.; Sumikama, T.; Chiba, J.; Ideguchi, E.; Saito, A.; Yamaguchi, T.; Hachiuma, I.; Suzuki, T.; Moriguchi, T.; Ozawa, A.; Ohtsubo, T.; Famiano, M. A.; Geissel, H.; Nettleton, A. S.; Tarasov, O. B.; Bazin, D.; Sherrill, B. M.; Manikonda, S. L.; Nolen, J. A.

    2012-11-01

    A search for isomeric ? decays among fission fragments from 345 MeV/nucleon 238U has been performed at the RIKEN Nishina Center RI Beam Factory. Fission fragments were selected and identified using the superconducting in-flight separator BigRIPS and were implanted in an aluminum stopper. Delayed ? rays were detected using three clover-type high-purity germanium detectors located at the focal plane within a time window of 20 ?s following the implantation. We identified a total of 54 microsecond isomers with half-lives of ˜0.1-10 ?s, including the discovery of 18 new isomers in very neutron-rich nuclei: 59Tim, 90Asm, 92Sem, 93Sem, 94Brm, 95Brm, 96Brm, 97Rbm, 108Nbm, 109Mom, 117Rum, 119Rum, 120Rhm, 122Rhm, 121Pdm, 124Pdm, 124Agm, and 126Agm, and obtained a wealth of spectroscopic information such as half-lives, ?-ray energies, ?-ray relative intensities, and ?? coincidences over a wide range of neutron-rich exotic nuclei. Proposed level schemes are presented for 59Tim, 82Gam, 92Brm, 94Brm, 95Brm, 97Rbm, 98Rbm, 108Nbm, 108Zrm, 109Mom, 117Rum, 119Rum, 120Rhm, 122Rhm, 121Pdm, 124Agm, and 125Agm, based on the obtained spectroscopic information and the systematics in neighboring nuclei. The nature of the nuclear isomerism is discussed in relation to the evolution of nuclear structure.

  15. Designing an upgrade of the Medley setup for light-ion production and fission cross-section measurements

    E-print Network

    Kaj Jansson; Cecilia Gustavsson; Ali Al-Adili; Anders Hjalmarsson; Erik Andersson-Sundén; Alexander V. Prokofiev; Diego Tarrío; Stephan Pomp

    2015-06-23

    Measurements of neutron-induced fission cross sections and light-ion production are planned in the energy range 1-40 MeV at the upcoming Neutrons For Science (NFS) facility. In order to prepare our detector setup for the neutron beam with continuous energy spectrum, a simulation software was written using the Geant4 toolkit for both measurement situations. The neutron energy range around 20 MeV is troublesome when it comes to the cross sections used by Geant4 since data-driven cross sections are only available below 20 MeV but not above, where they are based on semi-empirical models. Several customisations were made to the standard classes in Geant4 in order to produce consistent results over the whole simulated energy range. Expected uncertainties are reported for both types of measurements. The simulations have shown that a simultaneous precision measurement of the three standard cross sections H(n,n), $^{235}$U(n,f) and $^{238}$U(n,f) relative to each other is feasible using a triple layered target. As high resolution timing detectors for fission fragments we plan to use Parallel Plate Avalanche Counters (PPACs). The simulation results have put some restrictions on the design of these detectors as well as on the target design. This study suggests a fissile target no thicker than 2 micrometers (1.7 mg/cm$^2$) and a PPAC foil thickness preferably less than 1 micrometer. We also comment on the usability of Geant4 for simulation studies of neutron reactions in this energy range.

  16. Designing an upgrade of the Medley setup for light-ion production and fission cross-section measurements

    NASA Astrophysics Data System (ADS)

    Jansson, K.; Gustavsson, C.; Al-Adili, A.; Hjalmarsson, A.; Andersson-Sundén, E.; Prokofiev, A. V.; Tarrío, D.; Pomp, S.

    2015-09-01

    Measurements of neutron-induced fission cross-sections and light-ion production are planned in the energy range 1-40 MeV at the upcoming Neutrons For Science (NFS) facility. In order to prepare our detector setup for the neutron beam with continuous energy spectrum, a simulation software was written using the Geant4 toolkit for both measurement situations. The neutron energy range around 20 MeV is troublesome when it comes to the cross-sections used by Geant4 since data-driven cross-sections are only available below 20 MeV but not above, where they are based on semi-empirical models. Several customisations were made to the standard classes in Geant4 in order to produce consistent results over the whole simulated energy range. Expected uncertainties are reported for both types of measurements. The simulations have shown that a simultaneous precision measurement of the three standard cross-sections H(n,n), 235U(n,f) and 238U(n,f) relative to each other is feasible using a triple layered target. As high resolution timing detectors for fission fragments we plan to use Parallel Plate Avalanche Counters (PPACs). The simulation results have put some restrictions on the design of these detectors as well as on the target design. This study suggests a fissile target no thicker than 2 ?m (1.7 mg/cm2) and a PPAC foil thickness preferably less than 1 ?m. We also comment on the usability of Geant4 for simulation studies of neutron reactions in this energy range.

  17. Shale Gas Production: Potential versus Actual GHG Emissions

    E-print Network

    Shale Gas Production: Potential versus Actual GHG Emissions Francis O'Sullivan and Sergey Paltsev://globalchange.mit.edu/ Printed on recycled paper #12;1 Shale Gas Production: Potential versus Actual GHG Emissions Francis O'Sullivan* and Sergey Paltsev* Abstract Estimates of greenhouse gas (GHG) emissions from shale gas production and use

  18. Ionizing radiation accelerates Drp1-dependent mitochondrial fission, which involves delayed mitochondrial reactive oxygen species production in normal human fibroblast-like cells

    SciTech Connect

    Kobashigawa, Shinko; Suzuki, Keiji; Yamashita, Shunichi

    2011-11-04

    Highlights: Black-Right-Pointing-Pointer We report first time that ionizing radiation induces mitochondrial dynamic changes. Black-Right-Pointing-Pointer Radiation-induced mitochondrial fission was caused by Drp1 localization. Black-Right-Pointing-Pointer We found that radiation causes delayed ROS from mitochondria. Black-Right-Pointing-Pointer Down regulation of Drp1 rescued mitochondrial dysfunction after radiation exposure. -- Abstract: Ionizing radiation is known to increase intracellular level of reactive oxygen species (ROS) through mitochondrial dysfunction. Although it has been as a basis of radiation-induced genetic instability, the mechanism involving mitochondrial dysfunction remains unclear. Here we studied the dynamics of mitochondrial structure in normal human fibroblast like cells exposed to ionizing radiation. Delayed mitochondrial O{sub 2}{sup {center_dot}-} production was peaked 3 days after irradiation, which was coupled with accelerated mitochondrial fission. We found that radiation exposure accumulated dynamin-related protein 1 (Drp1) to mitochondria. Knocking down of Drp1 expression prevented radiation induced acceleration of mitochondrial fission. Furthermore, knockdown of Drp1 significantly suppressed delayed production of mitochondrial O{sub 2}{sup {center_dot}-}. Since the loss of mitochondrial membrane potential, which was induced by radiation was prevented in cells knocking down of Drp1 expression, indicating that the excessive mitochondrial fission was involved in delayed mitochondrial dysfunction after irradiation.

  19. Convergence Improvement and Qualification of a Model for Fission

    E-print Network

    Boye, Johan

    Convergence Improvement and Qualification of a Model for Fission Gas Behaviour in Nuclear Fuel G and Qualification of a Model for Fission Gas Behaviour in Nuclear Fuel G A Ë L D U B U S Master's Thesis OF A MODEL FOR FISSION GAS BEHAVIOUR IN NUCLEAR FUEL Abstract : During the irradiation in a fuel assembly

  20. The Fission Time Projection Chamber

    NASA Astrophysics Data System (ADS)

    Heffner, Mike

    2009-10-01

    New high-precision fission experiments have become a priority within the nuclear energy community due to a growing, world wide, interest in nuclear reactors. In particular, the designs of next generation reactors require reductions in the uncertainties on a number of energy dependent, neutron induced fission cross sections. The fission Time Projection Chamber (fission TPC) is the instrument that has been selected to carry out these challenging cross section measurements. This 6000 pad TPC with 2mm hex pads has a volume of only 2 liters and is filled with a hydrogen working gas. A unique set of electronics have been designed for the TPC that use all off the shelf components to reduce development costs. In this talk, I will show how the TPC will improve previous measurements, the design specifics of the fission TPC and the progress to date.

  1. Shale Gas Production: Potential versus Actual GHG Emissions

    E-print Network

    O'Sullivan, Francis

    Estimates of greenhouse gas (GHG) emissions from shale gas production and use are controversial. Here we assess the level of GHG emissions from shale gas well hydraulic fracturing operations in the United States during ...

  2. Isotope ratio analysis of actinides, fission products, and geolocators by high-efficiency multi-collector thermal ionization mass spectrometry

    NASA Astrophysics Data System (ADS)

    Bürger, S.; Riciputi, L. R.; Bostick, D. A.; Turgeon, S.; McBay, E. H.; Lavelle, M.

    2009-09-01

    A ThermoFisher "Triton" multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotope ratio analysis of actinides (uranium, plutonium, and americium), fission products and geolocators (strontium, cesium, and neodymium). Total efficiencies (atoms loaded to ions detected) of up to 0.5-2% for U, Pu, and Am, and 1-30% for Sr, Cs, and Nd can be reported employing resin bead load techniques onto flat ribbon Re filaments or resin beads loaded into a millimeter-sized cavity drilled into a Re rod. This results in detection limits of <0.1 fg (104 atoms to 105 atoms) for 239-242+244Pu, 233+236U, 241-243Am, 89,90Sr, and 134,135,137Cs, and <=1 pg for natural Nd isotopes (limited by the chemical processing blank) using a secondary electron multiplier (SEM) or multiple-ion counters (MICs). Relative standard deviations (RSD) as small as 0.1% and abundance sensitivities of 1 × 106 or better using a SEM are reported here. Precisions of RSD [approximate]0.01-0.001% using a multi-collector Faraday cup array can be achieved at sub-nanogram concentrations for strontium and neodymium and are suitable to gain crucial geolocation information. The analytical protocols reported herein are of particular value for nuclear forensic and nuclear safeguard applications.

  3. Electrolysis of uranium nitride containing fission product elements (Mo, Pd, Nd) in a molten LiCl-KCl eutectic

    SciTech Connect

    Satoh, Takumi; Iwai, Takashi; Arai, Yasuo

    2007-07-01

    The electrolysis of burnup-simulated uranium nitride, UN, containing representative solid fission product elements (Mo, Pd, Nd) was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl{sub 3} from the view point of application of pyrochemical reprocessing to nitride fuel cycle. It was found from cyclic voltammetry and anodic polarization curve measurement that anodic dissolution of UN began at about -0.75 V vs. Ag/AgCl reference electrode in all samples. After the electrolysis at the constant anodic potential of -0.65 {approx} -0.60 V vs. Ag/AgCl, most of UN was dissolved into LiCl- KCl as UCl{sub 3} at the anode, and U was recovered in the liquid Cd cathode in all samples. Further, Nd was dissolved into LiCl-KCl as NdCl{sub 3}, while Mo and Pd were not dissolved but remained at the anode. (authors)

  4. Wet deposition of fission-product isotopes to North America from the Fukushima Dai-ichi incident, March 2011

    USGS Publications Warehouse

    Wetherbee, Gregory A.; Gay, David A.; Debey, Timothy M.; Lehmann, Christopher M.B.; Nilles, Mark A.

    2012-01-01

    Using the infrastructure of the National Atmospheric Deposition Program (NADP), numerous measurements of radionuclide wet deposition over North America were made for 167 NADP sites before and after the Fukushima Dai-ichi Nuclear Power Station incident of March 12, 2011. For the period from March 8 through April 5, 2011, wet-only precipitation samples were collected by NADP and analyzed for fission-product isotopes within whole-water and filterable solid samples by the United States Geological Survey using gamma spectrometry. Variable amounts of 131I, 134Cs, or 137Cs were measured at approximately 21% of sampled NADP sites distributed widely across the contiguous United States and Alaska. Calculated 1- to 2-week individual radionuclide deposition fluxes ranged from 0.47 to 5100 Becquerels per square meter during the sampling period. Wet deposition activity was small compared to measured activity already present in U.S. soil. NADP networks responded to this complex disaster, and provided scientifically valid measurements that are comparable and complementary to other networks in North America and Europe.

  5. Designing an upgrade of the Medley setup for light-ion production and fission cross-section measurements

    E-print Network

    Jansson, Kaj; Al-Adili, Ali; Hjalmarsson, Anders; Andersson-Sundén, Erik; Prokofiev, Alexander V; Tarrío, Diego; Pomp, Stephan

    2015-01-01

    Measurements of neutron-induced fission cross sections and light-ion production are planned in the energy range 1-40 MeV at the upcoming Neutrons For Science (NFS) facility. In order to prepare our detector setup for the neutron beam with continuous energy spectrum, a simulation software was written using the Geant4 toolkit for both measurement situations. The neutron energy range around 20 MeV is troublesome when it comes to the cross sections used by Geant4 since data-driven cross sections are only available below 20 MeV but not above, where they are based on semi-empirical models. Several customisations were made to the standard classes in Geant4 in order to produce consistent results over the whole simulated energy range. Expected uncertainties are reported for both types of measurements. The simulations have shown that a simultaneous precision measurement of the three standard cross sections H(n,n), $^{235}$U(n,f) and $^{238}$U(n,f) relative to each other is feasible using a triple layered target. As hig...

  6. Release of fission products from irradiated SRP fuels at elevated temperature. Data report on the first stage of the SRP source term study

    SciTech Connect

    Woodley, R.E.

    1986-06-01

    For a sound evaluation of the consequences of a hypothetical nuclear reactor accident, a knowledge of the extent of fission product release from the fuel at anticipated temperatures and atmosphere conditions is required. Measurements of fission product release have been performed with a variety of nuclear fuels under various conditions of temperature and atmosphere. While the use of data obtained on fuels similar to the fuel of interest may provide a reasonable estimate of release fractions, precise information of this nature can only be obtained from measurements employing specimens of the actual fuels used in the nuclear reactor under consideration. The two fuels of interest in the present study are an alloy, a dispersion of UAl/sub 4/ in an aluminum matrix, and a cermet, a dispersion of U/sub 3/O/sub 8/ in an aluminum matrix. Both fuels are clad in aluminum.

  7. Benchmarking Nuclear Fission Theory

    E-print Network

    G. F. Bertsch; W. Loveland; W. Nazarewicz; P. Talou

    2015-02-20

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. The purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  8. Benchmarking nuclear fission theory

    SciTech Connect

    Bertsch, G. F.; Loveland, W.; Nazarewicz, W.; Talou, P.

    2015-05-14

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. Thus, the purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  9. Benchmarking Nuclear Fission Theory

    E-print Network

    Bertsch, G F; Nazarewicz, W; Talou, P

    2015-01-01

    We suggest a small set of fission observables to be used as test cases for validation of theoretical calculations. The purpose is to provide common data to facilitate the comparison of different fission theories and models. The proposed observables are chosen from fission barriers, spontaneous fission lifetimes, fission yield characteristics, and fission isomer excitation energies.

  10. Bio Gas Oil Production from Waste Lard

    PubMed Central

    Hancsók, Jen?; Baladincz, Péter; Kasza, Tamás; Kovács, Sándor; Tóth, Csaba; Varga, Zoltán

    2011-01-01

    Besides the second generations bio fuels, one of the most promising products is the bio gas oil, which is a high iso-paraffin containing fuel, which could be produced by the catalytic hydrogenation of different triglycerides. To broaden the feedstock of the bio gas oil the catalytic hydrogenation of waste lard over sulphided NiMo/Al2O3 catalyst, and as the second step, the isomerization of the produced normal paraffin rich mixture (intermediate product) over Pt/SAPO-11 catalyst was investigated. It was found that both the hydrogenation and the decarboxylation/decarbonylation oxygen removing reactions took place but their ratio depended on the process parameters (T = 280–380°C, P = 20–80 bar, LHSV = 0.75–3.0?h?1 and H2/lard ratio: 600?Nm3/m3). In case of the isomerization at the favourable process parameters (T = 360–370°C, P = 40 –50 bar, LHSV = 1.0?h?1 and H2/hydrocarbon ratio: 400?Nm3/m3) mainly mono-branching isoparaffins were obtained. The obtained products are excellent Diesel fuel blending components, which are practically free of heteroatoms. PMID:21403875

  11. CO Methanation for Synthetic Natural Gas Production.

    PubMed

    Kambolis, Anastasios; Schildhauer, Tilman J; Kröcher, Oliver

    2015-01-01

    Energy from woody biomass could supplement renewable energy production towards the replacement of fossil fuels. A multi-stage process involving gasification of wood and then catalytic transformation of the producer gas to synthetic natural gas (SNG) represents progress in this direction. SNG can be transported and distributed through the existing pipeline grid, which is advantageous from an economical point of view. Therefore, CO methanation is attracting a great deal of attention and much research effort is focusing on the understanding of the process steps and its further development. This short review summarizes recent efforts at Paul Scherrer Institute on the understanding of the reaction mechanism, the catalyst deactivation, and the development of catalytic materials with benign properties for CO methanation. PMID:26598405

  12. The use of WIMS-ANL lumped fission product cross sections for burned core analysis with the MCNP Monte Carlo code.

    SciTech Connect

    Hanan, N. A.

    1998-10-14

    Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel assemblies, with LEU and HEU fuels, are provided to demonstrate the use of this method.

  13. Evaluation of containment failure modes and fission product releases during core meltdown accidents in a BWR with a Mark III containment

    SciTech Connect

    Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

    1985-01-01

    An assessment is described of potential failure modes and fission product releases for a large number of postulated core meltdown accidents in a BWR with a Mark III containment. For this containment design, the most important failure mode was found to be due to hydrogen related phenomena. A one-dimensional lumped parameter computer code has been developed and used to determine the probability of various hydrogen phenomena for a range of postulated core meltdown sequences. Potential containment loads have been estimated and compared against the containment capacity to determine the probability of containment failure. The fission product release assessment began by using the MARCH/CORRAL system of codes with key input parameters varied over a reasonable range. The parameters relate to primary system retention, re-emission, pool scrubbing, and fission product release in-vessel vs ex-vessel. The final step used more mechanistic calculations based on the system of codes recently developed under sponsorship of the Accident Source Term Program Office, NRC, and compares these predictions with the range of releases calculated in the sensitivity study.

  14. Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

    NASA Astrophysics Data System (ADS)

    Sakama, M.; Nagano, Y.; Kitade, T.; Shikino, O.; Nakayama, S.

    2014-06-01

    Radioactive fission product 131I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, 134Cs and 137Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m-3 in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of 134Cs and 137Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m-3) variation of stable cesium (133Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

  15. Bimodal fission

    SciTech Connect

    Hulet, E.K.

    1989-04-19

    In recent years, we have measured the mass and kinetic-energy distributions from the spontaneous fission of /sup 258/Fm, /sup 259/Md, /sup 260/Md, /sup 258/No, /sup 262/No, and /sup 260/(104). All are observed to fission with a symmetrical division of mass, whereas the total-kinetic-energy (TKE) distributions strongly deviated from the Gaussian shape characteristically found in the fission of all other actinides. When the TKE distributions are resolved into two Gaussians the constituent peaks lie near 200 and near 233 MeV. We conclude two modes or bimodal fission is occurring in five of the six nuclides studied. Both modes are possible in the same nuclides, but one generally predominates. We also conclude the low-energy but mass-symmetrical mode is likely to extend to far heavier nuclei; while the high-energy mode will be restricted to a smaller region, a region of nuclei defined by the proximity of the fragments to the strong neutron and proton shells in /sup 132/Sn. 16 refs., 7 figs., 1 tab.

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

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

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

    SciTech Connect

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

    2010-02-26

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  18. Fission dynamics within time-dependent Hartree-Fock: deformation-induced fission

    E-print Network

    Goddard, P M; Rios, A

    2015-01-01

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe fast fission processes beyond the fission barrier, using the nuclide $^{240}$Pu as an example. Methods: Time-dependent Hartree-Fock calculations based on the Skyrme interaction are used to calculate non-adiabatic fission paths, beginning from static constrained Hartree-Fock calculations. The properties of the dynamic states are interpreted in terms of the nature of their collective motion. Fission product properties are compared to data. Results: Parent nuclei constrained to begin dynamic evolution with a deformation less than the fission barrier exhibit giant-resonance-type behaviour. Those beginning just beyond the ...

  19. Uncertainty Analysis on Fission Molybdenum Production with a Nuclear Fuel Target in a Research Reactor

    SciTech Connect

    Cho, Dong-Keun; Kim, Myung-Hyun

    2003-10-15

    The use of a low-enriched uranium (LEU) fuel target was examined for the feasibility of {sup 99}Mo production in a High-flux Advanced Neutron Application Reactor (HANARO). Uncertainty analysis was done with respect to the {sup 99}Mo yield ratio, {sup 239}Pu yield ratio, annual production rate, and decontamination requirement. Validity of a coupled code system, MCNP/ORIGEN2, was evaluated to estimate reliable isotopic number densities after irradiation and cooling. An equilibrium core model for the MCNP fixed-source problem was found by the reactor design methodology known as WIMS/VENTURE. Optimized target design options were proposed for both the LEU and highly enriched uranium (HEU) targets. Variables related to the target fabrication process and reactor physics condition were considered as uncertainty-inducing parameters. The most important factor affecting the overall uncertainty of the LEU option was the engineering tolerances achievable in the fabrication process of fuel film. The LEU has twice the uncertainty of HEU under current technology, which makes the economics of LEU worse than HEU. It is acceptable, however, in view of the radioactive purity of the alpha emitter because the uncertainty of the impurity level of {sup 239}Pu is expected to be relatively small - only 6.5% with a 95% confidence level.

  20. Mitochondrial fission induces glycolytic reprogramming in cancer-associated myofibroblasts, driving stromal lactate production, and early tumor growth.

    PubMed

    Guido, Carmela; Whitaker-Menezes, Diana; Lin, Zhao; Pestell, Richard G; Howell, Anthony; Zimmers, Teresa A; Casimiro, Mathew C; Aquila, Saveria; Ando', Sebastiano; Martinez-Outschoorn, Ubaldo E; Sotgia, Federica; Lisanti, Michael P

    2012-08-01

    Recent studies have suggested that cancer cells behave as metabolic parasites, by inducing oxidative stress in adjacent normal fibroblasts. More specifically, oncogenic mutations in cancer cells lead to ROS production and the "secretion" of hydrogen peroxide species. Oxidative stress in stromal fibroblasts then induces their metabolic conversion into cancer-associated fibroblasts. Such oxidative stress drives the onset of autophagy, mitophagy, and aerobic glycolysis in fibroblasts, resulting in the local production of high-energy mitochondrial fuels (such as L-lactate, ketone bodies, and glutamine). These recycled nutrients are then transferred to cancer cells, where they are efficiently burned via oxidative mitochondrial metabolism (OXPHOS). We have termed this new energy-transfer mechanism "Two-Compartment Tumor Metabolism", to reflect that the production and consumption of nutrients (L-lactate and other catabolites) is highly compartmentalized. Thus, high-energy onco-catabolites are produced by the tumor stroma. Here, we used a genetic approach to stringently test this energy-transfer hypothesis. First, we generated hTERT-immortalized fibroblasts which were genetically re-programmed towards catabolic metabolism. Metabolic re-programming towards glycolytic metabolism was achieved by the recombinant over-expression of MFF (mitochondrial fission factor). MFF over-expression results in extensive mitochondrial fragmentation, driving mitochondrial dysfunction. Our results directly show that MFFfibroblasts undergo oxidative stress, with increased ROS production, and the onset of autophagy and mitophagy, both catabolic processes. Mechanistically, oxidative stress induces autophagy via NF-kB activation, also providing a link with inflammation. As a consequence MFF-fibroblasts showed intracellular ATP depletion and the extracellular secretion of L-lactate, a critical onco-catabolite. MFF-fibroblasts also showed signs of myofibroblast differentiation, with the expression of SMA and calponin. Importantly, MFF-fibroblasts signficantly promoted early tumor growth (up to 6.5-fold), despite a 20% overall reduction in angiogenesis. Thus, catabolic metabolism in cancer-associated fibroblasts may be a critical event during tumor intiation, allowing accelerated tumor growth, especially prior to the onset of neoangiogenesis. PMID:22878233

  1. Effects of gas bubble production on heat transfer from a volumetrically heated liquid pool

    NASA Astrophysics Data System (ADS)

    Bull, Geoffrey R.

    Aqueous solutions of uranium salts may provide a new supply chain to fill potential shortfalls in the availability of the most common radiopharmaceuticals currently in use worldwide, including Tc99m which is a decay product of Mo99. The fissioning of the uranium in these solutions creates Mo99 but also generates large amounts of hydrogen and oxygen from the radiolysis of the water. When the dissolved gases reach a critical concentration, bubbles will form in the solution. Bubbles in the solution affect both the fission power and the heat transfer out of the solution. As a result, for safety and production calculations, the effects of the bubbles on heat transfer must be understood. A high aspect ratio tank was constructed to simulate a section of an annulus with heat exchangers on the inner and outer steel walls to provide cooling. Temperature measurements via thermocouples inside the tank and along the outside of the steel walls allowed the calculation of overall and local heat transfer coefficients. Different air injection manifolds allowed the exploration of various bubble characteristics and patterns on heat transfer from the pool. The manifold type did not appear to have significant impact on the bubble size distributions in water. However, air injected into solutions of magnesium sulfate resulted in smaller bubble sizes and larger void fractions than those in water at the same injection rates. One dimensional calculations provide heat transfer coefficient values as functions of the superficial gas velocity in the pool.

  2. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  3. Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

    SciTech Connect

    Rimshaw, S.J.; Case, F.N.

    1980-01-01

    Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO/sub 3/ ..-->.. N/sub 2/O + 4 CO/sub 2/ + 5 H/sub 2/O. These gases can be scrubbed free of /sup 106/Ru, /sup 131/I, and /sup 99/Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO/sub 4/ volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO/sub 4/ in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO/sub 2/ aerosol having a particle size of 0.5 ..mu... This RuO/sub 2/ aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250/sup 0/C and a moderate volatility of 9.5% at 600/sup 0/C. Volatilities of /sup 99/Tc were generally <1% on calcining acidic or basic wastes at temperatures of 250 to 600/sup 0/C. Data are presented to indicate that /sup 99/Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg /sup 99/Tc being associated with each curie of /sup 137/Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste.

  4. Fifty years with nuclear fission

    SciTech Connect

    Behrens, J.W.; Carlson, A.D. )

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  5. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    SciTech Connect

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling methods used in this study.

  6. Ground movements associated with gas hydrate production

    SciTech Connect

    Siriwardane, H.J.

    1992-10-01

    The mechanics of ground movements during hydrate production can be more closely simulated by considering similarities with ground movements associated with subsidence in permafrost regions than with gob compaction in a longwall mine. The purpose of this research work is to investigate the potential strata movements associated with hydrate production by considering similarities with ground movements in permafrost regions. The work primarily involves numerical modeling of subsidence caused by hydrate dissociation. The investigation is based on the theories of continuum mechanics , thermomechanical behavior of frozen geo-materials, and principles of rock mechanics and geomechanics. It is expected that some phases of the investigation will involve the use of finite element method, which is a powerful computer-based method which has been widely used in many areas of science and engineering. Parametric studies will be performed to predict expected strata movements and surface subsidence for different reservoir conditions and properties of geological materials. The results from this investigation will be useful in predicting the magnitude of the subsidence problem associated with gas hydrate production. The analogy of subsidence in permafrost regions may provide lower bounds for subsidence expected in hydrate reservoirs. Furthermore, it is anticipated that the results will provide insight into planning of hydrate recovery operations.

  7. Measurements of Methane Emissions at Natural Gas Production Sites

    E-print Network

    Lightsey, Glenn

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

  8. Microbial Dynamics and Control in Shale Gas Production Jason Gaspar,

    E-print Network

    Alvarez, Pedro J.

    Microbial Dynamics and Control in Shale Gas Production Jason Gaspar, Jacques Mathieu, Yu Yang, Ross effects in shale gas production, such as reservoir souring, plugging, equipment corrosion, and a decrease fluids, drilling mud, and impoundment water likely introduce deleterious microorganisms into shale gas

  9. Fission meter

    DOEpatents

    Rowland, Mark S. (Alamo, CA); Snyderman, Neal J. (Berkeley, CA)

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  10. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel experienced high linear powers during the first year of irradiation, but at the end of the IFA-609/626 period, the average linear power of the rod was around 12 kW/m. In the IFA-702, the power was gradually increased over 7 days from 12 kW/m to 22.5 kW/m before it was decreased again to reach ˜19 kW/m at the end of the 100 days forming this part of the irradiation. A LEICA (DM RXA2) optical microscope. A shielded electronic microprobe (EPMA) SX-100R by CAMECA. A shielded scanning electron microscope (SEM): the Philips XL30. Image acquisitions were performed using the ADDA "SIS" system with the AnalySIS software for image analysis. A shielded secondary ion mass spectrometer (SIMS): the CAMECA IMS 6f was capable of analysing the same samples as the SEM and EPMA [16-22]. In the central part of the pellet for all three phases, Xe precipitated into bubbles with very little Xe remaining outside the bubbles. Some Xe-filled bubbles were detected under the surface in this area. They appear as bright spots. Around mid-radius on the periphery of the 235U-poor areas and in the intermediate phase, Xe was depleted on the periphery of the grains. This depletion was not associated with Xe-filled bubbles that would be detected under the polished surface. Moreover, no large intergranular open bubbles were visible. Therefore, this missing gas must have been released. In the 235U-rich agglomerates all over the section, Xe precipitated into bubbles with very little Xe remaining outside the bubbles. The Xe quantitative analyses through 235U-rich agglomerates on the pellet periphery (Fig. 9) confirmed the low quantity of Xe remaining outside the bubbles. This Xe content was around 0.1 wt%. Fig. 10 shows the Xe and Nd EPMA quantitative measurements along a radius of the cross section. In this figure and in Fig. 9, the weight percentage scales were set so that the two profiles would be almost identical without Xe release or precipitation. Along the Xe axis, the Nd profile can be considered as the local Xe production. Fig. 10 shows that the Xe measurement all through the

  11. Microscopic description of complex nuclear decay: Multimodal fission

    SciTech Connect

    Staszczak, A.; Baran, A.; Dobaczewski, J.; Nazarewicz, W.

    2009-07-15

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  12. Microscopic description of complex nuclear decay: multimodal fission

    E-print Network

    Staszczak, A; Dobaczewski, J; Nazarewicz, W

    2009-01-01

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  13. Microscopic description of complex nuclear decay: Multimodal fission

    NASA Astrophysics Data System (ADS)

    Staszczak, A.; Baran, A.; Dobaczewski, J.; Nazarewicz, W.

    2009-07-01

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  14. Microscopic description of complex nuclear decay: multimodal fission

    E-print Network

    A. Staszczak; A. Baran; J. Dobaczewski; W. Nazarewicz

    2009-06-23

    Our understanding of nuclear fission, a fundamental nuclear decay, is still incomplete due to the complexity of the process. In this paper, we describe a study of spontaneous fission using the symmetry-unrestricted nuclear density functional theory. Our results show that the observed bimodal fission can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. We also predict a new phenomenon of trimodal spontaneous fission for some rutherfordium, seaborgium, and hassium isotopes.

  15. Background radiation from fission pulses

    SciTech Connect

    England, T.R.; Arthur, E.D.; Brady, M.C.; LaBauve, R.J.

    1988-05-01

    Extensive source terms for beta, gamma, and neutrons following fission pulses are presented in various tabular and graphical forms. Neutron results from a wide range of fissioning nuclides (42) are examined and detailed information is provided for four fuels: /sup 235/U, /sup 238/U, /sup 232/Th, and /sup 239/Pu; these bracket the range of the delayed spectra. Results at several cooling (decay) times are presented. For ..beta../sup -/ and ..gamma.. spectra, only /sup 235/U and /sup 239/Pu results are given; fission-product data are currently inadequate for other fuels. The data base consists of all known measured data for individual fission products extensively supplemented with nuclear model results. The process is evolutionary, and therefore, the current base is summarized in sufficient detail for users to judge its quality. Comparisons with recent delayed neutron experiments and total ..beta../sup -/ and ..gamma.. decay energies are included. 27 refs., 47 figs., 9 tabs.

  16. Saturation current of miniaturized fission chambers

    NASA Astrophysics Data System (ADS)

    Chabod, Sébastien P.

    2009-01-01

    We present the detailed formulae of the saturation currents for the four main categories of fission chambers operating in current mode. The results obtained are function of simple parameters: number of fission reactions within the chamber deposits, geometric characteristics of the electrodes and filling gas properties. A direct relation between the saturation current values and the ambient neutron flux is thus established. These results should reduce the number, the duration and the cost of the calibration procedures required to operate the fission chambers.

  17. Fission dynamics within time-dependent Hartree-Fock: boost-induced fission

    E-print Network

    P. M. Goddard; P. D. Stevenson; A. Rios

    2015-10-27

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe induced fission processes, using quadrupole boosts in the nuclide $^{240}$Pu as an example. Methods: Quadrupole constrained Hartree-Fock calculations are used to create a potential energy surface. An isomeric state and a state beyond the second barrier peak are excited by means of instantaneous as well as temporally extended gauge boosts with quadrupole shapes. The subsequent deexcitation is studied in a time-dependent Hartree-Fock simulation, with emphasis on fissioned final states. The corresponding fission fragment mass numbers are studied. Results: In general, the energy deposited by the quadrupole boost is quickly absorbed by the nucleus. In instantaneous boosts, this leads to fast shape rearrangements and violent dynamics that can ultimately lead to fission. This is a qualitatively different process than the deformation-induced fission. Boosts induced within a finite time window excite the system in a relatively gentler way, and do induce fission but with a smaller energy deposition. Conclusions: The fission products obtained using boost-induced fission in time-dependent Hartree-Fock are more asymmetric than the fragments obtained in deformation-induced fission, or the corresponding adiabatic approaches.

  18. Optimization and Evaluation of Mixed-Bed Chemisorbents for Extracting Fission and Activation Products from Marine and Fresh Waters

    SciTech Connect

    Johnson, Bryce; Santschi, Peter H.; Addleman, Raymond S.; Douglas, Matthew; Davidson, Joseph D.; Fryxell, Glen E.; Schwantes, Jon M.

    2011-06-02

    Chemically selective chemisorbents are needed to monitor natural and engineered waters for anthropogenic releases of stable and radioactive contaminants. Here, a number of individual and mixtures of chemisorbents were investigated for their ability to extract select fission and activation product elements from marine and coastal waters, including Co, Zr, Ru, Ag, Te, Sb, Ba, Cs, Ce, Eu, Pa, Np, and Th. Conventional manganese oxide and cyanoferrate sorbents, including commercially available Anfezh and potassium hexacyanocobalt(II) ferrate(II) (KCFC), were tested along with novel nano-structured surfaces (known as Self Assembled Monolayers on Mesoporous Supports or SAMMS) functionalized with a variety of moieties including thiol, diphosphonic acid (DiPhos-), methyl, 3, 4 hydroxypyridinone (HOPO-), and cyanoferrate. Extraction efficiencies were measured as a function of salinity, organic content, temperature, flow rate and sample size for both synthetic and natural fresh and saline waters under a range of environmentally relevant conditions. The effect of flow rate on extraction efficiency, from 1 to 70 mL min-1, provided some insight on rate limitations of mechanisms affecting sorption processes. Optimized mixtures of sorbent-ligand chemistries afforded excellent retention of all target elements, except, Ba and Sb. Mixtures of tested chemisorbents, including MnO2/Anfezh and MnO2/KCFC/Thiol (1-3mm)-SAMMS, extracted 8 of the 11 target elements studied to better than 80% efficiency, while a mixture of MnO2/Anfezh/Thiol (75-150 {mu}m)-SAMMS mixture was able to extract 7 of the 11 target elements to better than 90%. Results generated here indicate that flow rate should be less of a consideration for experimental design if sampling from fresh water containing variable amounts of DOM, rather than collecting samples from salt water environments. Relative to the capability of any single type of chemisorbent tested, optimized mixtures of several sorbents are able to increase the number of elements that can be efficiently and simultaneously extracted from natural waters.

  19. Gas Production Strategy of Underground Coal Gasification Based on Multiple Gas Sources

    PubMed Central

    Tianhong, Duan; Zuotang, Wang; Limin, Zhou; Dongdong, Li

    2014-01-01

    To lower stability requirement of gas production in UCG (underground coal gasification), create better space and opportunities of development for UCG, an emerging sunrise industry, in its initial stage, and reduce the emission of blast furnace gas, converter gas, and coke oven gas, this paper, for the first time, puts forward a new mode of utilization of multiple gas sources mainly including ground gasifier gas, UCG gas, blast furnace gas, converter gas, and coke oven gas and the new mode was demonstrated by field tests. According to the field tests, the existing power generation technology can fully adapt to situation of high hydrogen, low calorific value, and gas output fluctuation in the gas production in UCG in multiple-gas-sources power generation; there are large fluctuations and air can serve as a gasifying agent; the gas production of UCG in the mode of both power and methanol based on multiple gas sources has a strict requirement for stability. It was demonstrated by the field tests that the fluctuations in gas production in UCG can be well monitored through a quality control chart method. PMID:25114953

  20. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (inventors)

    1982-01-01

    A thermal reactor apparatus and method of pyrolyticaly decomposing silane gas into liquid silicon product and hydrogen by-product gas is disclosed. The thermal reactor has a reaction chamber which is heated well above the decomposition temperature of silane. An injector probe introduces the silane gas tangentially into the reaction chamber to form a first, outer, forwardly moving vortex containing the liquid silicon product and a second, inner, rewardly moving vortex containing the by-product hydrogen gas. The liquid silicon in the first outer vortex deposits onto the interior walls of the reaction chamber to form an equilibrium skull layer which flows to the forward or bottom end of the reaction chamber where it is removed. The by-product hydrogen gas in the second inner vortex is removed from the top or rear of the reaction chamber by a vortex finder. The injector probe which introduces the silane gas into the reaction chamber is continually cooled by a cooling jacket.

  1. A suitability study of the fission product phantom and the bottle manikin absorption phantom for calibration of in vivo bioassay equipment for the DOELAP accreditation testing program

    SciTech Connect

    Olsen, P.C.; Lynch, T.P.

    1991-08-01

    Pacific Northwest laboratory (PNL) conducted an intercomparison study of the Fission Product phantom and the bottle manikin absorption (BOMAB) phantom for the US Department of Energy (DOE) to determine the consistency of calibration response of the two phantoms and their suitability for certification and use under a planned bioassay laboratory accreditation program. The study was initiated to determine calibration factors for both types of phantoms and to evaluate the suitability of their use in DOE Laboratory Accreditation Program (DOELAP) round-robin testing. The BOMAB was found to be more appropriate for the DOELAP testing program. 9 refs., 9 figs., 9 tabs.

  2. On the fission chamber pulse charge acquisition and interpretation at MINERVE

    NASA Astrophysics Data System (ADS)

    Loiseau, P.; Geslot, B.; André, J.

    2013-04-01

    Fission Chambers (FCs) are widely used as neutron detectors for online flux measurement. The FC current pulse charge is a key observable quantity which depends on specifications such as the filling gas pressure and the FC geometry. In order to study pulse charges, experimental data have been acquired at the Cadarache zero power reactor MINERVE. Two chambers with contrasting specifications have been used. The experimental pulse charge spectrum is interpreted by the mean of a modeling of fission products (FPs) energy deposition within the filling gas. The pulse charge spectrum peaks are found to correspond to FP emitted perpendicularly to the electrodes.

  3. First-principles study of fission product (Xe, Cs, Sr) incorporation and segregation in alkaline earth metal oxides, HfO(2), and the MgO-HfO(2) interface.

    PubMed

    Liu, Xiang-Yang; Uberuaga, Blas P; Sickafus, Kurt E

    2009-01-28

    In order to close the nuclear fuel cycle, advanced concepts for separating out fission products are necessary. One approach is to use a dispersion fuel form in which a fissile core is surrounded by an inert matrix that captures and immobilizes the fission products from the core. If this inert matrix can be easily separated from the fuel, via e.g. solution chemistry, the fission products can be separated from the fissile material. We examine a surrogate dispersion fuel composition, in which hafnia (HfO(2)) is a surrogate for the fissile core and alkaline earth metal oxides are used as the inert matrix. The questions of fission product incorporation in these oxides and possible segregation behavior at interfaces are considered. Density functional theory based calculations for fission product elements (Xe, Sr, and Cs) in these oxides are carried out. We find smaller incorporation energy in hafnia than in MgO for Cs and Sr, and Xe if variation of charge state is allowed. We also find that this trend is reversed or reduced for alkaline earth metal oxides with large cation sizes. Model interfacial calculations show a strong tendency of segregation from bulk MgO to MgO-HfO(2) interfaces. PMID:21715804

  4. GASCAP: Wellhead Gas Productive Capacity Model documentation, June 1993

    SciTech Connect

    Not Available

    1993-07-01

    The Wellhead Gas Productive Capacity Model (GASCAP) has been developed by EIA to provide a historical analysis of the monthly productive capacity of natural gas at the wellhead and a projection of monthly capacity for 2 years into the future. The impact of drilling, oil and gas price assumptions, and demand on gas productive capacity are examined. Both gas-well gas and oil-well gas are included. Oil-well gas productive capacity is estimated separately and then combined with the gas-well gas productive capacity. This documentation report provides a general overview of the GASCAP Model, describes the underlying data base, provides technical descriptions of the component models, diagrams the system and subsystem flow, describes the equations, and provides definitions and sources of all variables used in the system. This documentation report is provided to enable users of EIA projections generated by GASCAP to understand the underlying procedures used and to replicate the models and solutions. This report should be of particular interest to those in the Congress, Federal and State agencies, industry, and the academic community, who are concerned with the future availability of natural gas.

  5. Analysis of eastern Devonian gas shales production data

    SciTech Connect

    Gatens, J.M.; Stanley, D.K.; Lancaster, D.E.; Lee, W.J.; Lane, H.S.; Watson, A.T.

    1989-05-01

    Production data from more than 800 Devonian shale wells have been analyzed. Permeability-thickness product and gas in place estimated from production data have been found to correlate with well performance. Empirical performance equations, production type curves, and an analytical dual-porosity model with automatic history-matching scheme were developed for the Devonian shale.

  6. Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

    SciTech Connect

    Hsieh, T.C.; Billone, M.C.; Rest, J.

    1982-03-01

    The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects.

  7. Forecasting long-term gas production Luis Cueto-Felguerosoa

    E-print Network

    Patzek, Tadeusz W.

    by increasing the length of a single well within the gas-bearing shale. Hydraulic fracturing, or "fracking" (9COMMENTARY Forecasting long-term gas production from shale Luis Cueto-Felguerosoa and Ruben Juanesa, Massachusetts Institute of Technology, Cambridge, MA 02139 Oil and natural gas from deep shale forma- tions

  8. Advanced Space Fission Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Borowski, Stanley K.

    2010-01-01

    Fission has been considered for in-space propulsion since the 1940s. Nuclear Thermal Propulsion (NTP) systems underwent extensive development from 1955-1973, completing 20 full power ground tests and achieving specific impulses nearly twice that of the best chemical propulsion systems. Space fission power systems (which may eventually enable Nuclear Electric Propulsion) have been flown in space by both the United States and the Former Soviet Union. Fission is the most developed and understood of the nuclear propulsion options (e.g. fission, fusion, antimatter, etc.), and fission has enjoyed tremendous terrestrial success for nearly 7 decades. Current space nuclear research and technology efforts are focused on devising and developing first generation systems that are safe, reliable and affordable. For propulsion, the focus is on nuclear thermal rockets that build on technologies and systems developed and tested under the Rover/NERVA and related programs from the Apollo era. NTP Affordability is achieved through use of previously developed fuels and materials, modern analytical techniques and test strategies, and development of a small engine for ground and flight technology demonstration. Initial NTP systems will be capable of achieving an Isp of 900 s at a relatively high thrust-to-weight ratio. The development and use of first generation space fission power and propulsion systems will provide new, game changing capabilities for NASA. In addition, development and use of these systems will provide the foundation for developing extremely advanced power and propulsion systems capable of routinely and affordably accessing any point in the solar system. The energy density of fissile fuel (8 x 10(exp 13) Joules/kg) is more than adequate for enabling extensive exploration and utilization of the solar system. For space fission propulsion systems, the key is converting the virtually unlimited energy of fission into thrust at the desired specific impulse and thrust-to-weight ratio. This presentation will discuss potential space fission propulsion options ranging from first generation systems to highly advanced systems. Ongoing research that shows promise for enabling second generation NTP systems with Isp greater than 1000 s will be discussed, as will the potential for liquid, gas, or plasma core systems. Space fission propulsion systems could also be used in conjunction with simple (water-based) propellant depots to enable routine, affordable missions to various destinations (e.g. moon, Mars, asteroids) once in-space infrastructure is sufficiently developed. As fuel and material technologies advance, very high performance Nuclear Electric Propulsion (NEP) systems may also become viable. These systems could enable sophisticated science missions, highly efficient cargo delivery, and human missions to numerous destinations. Commonalities between NTP, fission power systems, and NEP will be discussed.

  9. Fission dynamics within time-dependent Hartree-Fock: Deformation-induced fission

    NASA Astrophysics Data System (ADS)

    Goddard, Philip; Stevenson, Paul; Rios, Arnau

    2015-11-01

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus and the daughter products. Purpose: We explore the ability of dynamic mean-field methods to describe fast fission processes beyond the fission barrier, using the nuclide Pu240 as an example. Methods: Time-dependent Hartree-Fock calculations based on the Skyrme interaction are used to calculate nonadiabatic fission paths, beginning from static constrained Hartree-Fock calculations. The properties of the dynamic states are interpreted in terms of the nature of their collective motion. Fission product properties are compared to data. Results: Parent nuclei constrained to begin dynamic evolution with a deformation less than the fission barrier exhibit giant-resonance-type behavior. Those beginning just beyond the barrier explore large-amplitude motion but do not fission, whereas those beginning beyond the two-fragment pathway crossing fission to final states which differ according to the exact initial deformation. Conclusions: Time-dependent Hartree-Fock is able to give a good qualitative and quantitative description of fast fission, provided one begins from a sufficiently deformed state.

  10. Fission dynamics within time-dependent Hartree-Fock: deformation-induced fission

    E-print Network

    P. M. Goddard; P. D. Stevenson; A. Rios

    2015-11-03

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe fast fission processes beyond the fission barrier, using the nuclide $^{240}$Pu as an example. Methods: Time-dependent Hartree-Fock calculations based on the Skyrme interaction are used to calculate non-adiabatic fission paths, beginning from static constrained Hartree-Fock calculations. The properties of the dynamic states are interpreted in terms of the nature of their collective motion. Fission product properties are compared to data. Results: Parent nuclei constrained to begin dynamic evolution with a deformation less than the fission barrier exhibit giant-resonance-type behaviour. Those beginning just beyond the barrier explore large amplitude motion but do not fission, whereas those beginning beyond the two-fragment pathway crossing fission to final states which differ according to the exact initial deformation. Conclusions: Time-dependent Hartree-Fock is able to give a good qualitative and quantitative description of fast fission, provided one begins from a sufficiently deformed state.

  11. Diversification of 99Mo/99mTc separation: non–fission reactor production of 99Mo as a strategy for enhancing 99mTc availability.

    PubMed

    Pillai, Maroor R A; Dash, Ashutosh; Knapp, Furn F Russ

    2015-01-01

    This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,?) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo. PMID:25537991

  12. Measuring micro-organism gas production

    NASA Technical Reports Server (NTRS)

    Wilkins, J. R.; Pearson, A. O.; Mills, S. M.

    1973-01-01

    Transducer, which senses pressure buildup, is easy to assemble and use, and rate of gas produced can be measured automatically and accurately. Method can be used in research, in clinical laboratories, and for environmental pollution studies because of its ability to detect and quantify rapidly the number of gas-producing microorganisms in water, beverages, and clinical samples.

  13. Bulk-nanocrystalline oxide nuclear fuels - An innovative material option for increasing fission gas retention, plasticity and radiation-tolerance

    NASA Astrophysics Data System (ADS)

    Spino, J.; Santa Cruz, H.; Jovani-Abril, R.; Birtcher, R.; Ferrero, C.

    2012-03-01

    Advantages and disadvantages of bulk nanocrystalline (nc)-oxides (UO2, ZrO2, ThO2) and suggestions for their potential use as nuclear fuels and inert matrix carriers are described in this work on the basis of a study with nc-4 mol% Y2O3-ZrO2 bodies, which are envisaged to behave akin to highly exposed LWR-fuels with the High Burn-up Structure (HBS) also known as rim transformation. The main attributes of nc-fuels in-pile compared to conventional fuels will be the capacity to develop closed porosity retaining most of the fission gases, the ability to relax more efficiently the interaction stresses with the cladding (through much higher plasticity) and the enhanced resistance against radiation-damage thanks to their nanostructure. The present analysis comprises the long-term thermal stability of a porous nc-material, its property vs. porosity relations, the topology of the pore phase via X-ray synchrotron tomography, the behaviour under compressive stress and the performance under intense Xe-ions irradiation. Salient outcomes are the non-connectivity of the pore phase, the superplasticity of the nc-bodies and their high radiation-amorphisation resistance with negligible swelling under Xe-bombardment. Another important outcome of the present study is that deterioration of the thermal properties due to grain boundary effects (Kapitza resistance, melting point depression) can likely be avoided if the grain size is kept above 100 nm and, emulating the real HBS material, preferably in the range between 200 and 300 nm.

  14. Integrated production of fuel gas and oxygenated organic compounds from synthesis gas

    DOEpatents

    Moore, Robert B. (Allentown, PA); Hegarty, William P. (State College, PA); Studer, David W. (Wescosville, PA); Tirados, Edward J. (Easton, PA)

    1995-01-01

    An oxygenated organic liquid product and a fuel gas are produced from a portion of synthesis gas comprising hydrogen, carbon monoxide, carbon dioxide, and sulfur-containing compounds in a integrated feed treatment and catalytic reaction system. To prevent catalyst poisoning, the sulfur-containing compounds in the reactor feed are absorbed in a liquid comprising the reactor product, and the resulting sulfur-containing liquid is regenerated by stripping with untreated synthesis gas from the reactor. Stripping offgas is combined with the remaining synthesis gas to provide a fuel gas product. A portion of the regenerated liquid is used as makeup to the absorber and the remainder is withdrawn as a liquid product. The method is particularly useful for integration with a combined cycle coal gasification system utilizing a gas turbine for electric power generation.

  15. Measurement and calculation of the efficiency of fission detectors designed to monitor the time dependence of the neutron production of JET

    NASA Astrophysics Data System (ADS)

    Swinhoe, M. T.; Jarvis, O. N.

    1985-05-01

    Three pairs of fission counters (each pair one 235U and one 238U) are used at the Joint European Torus to determine the time dependence of the neutron production. In order to determine the absolute value of the neutron flux at the detector location it is necessary to know the neutron detection efficiency of the counter assemblies. This was measured using monoenergetic neutrons (at 2.5 and 14 MeV) and Cf and Am/Be sources. The fraction of fissions detected was determined by extrapolation of the pulse-height spectrum to zero pulse height. The calculation of efficiency was made with the Monte-Carlo neutron transport code MORSE. It was found that the detailed structure of the counter significantly affected the calculated efficiency and that the thermal cross-section values of the DLC37F nuclear data library had to be replaced with room-temperature values. The mean difference between calculation and experiment is (5.5±6.3)%.

  16. Methane hydrate gas production: evaluating and exploiting the solid gas resource

    SciTech Connect

    McGuire, P.L.

    1981-01-01

    Methane hydrate gas could be a tremendous energy resource if methods can be devised to produce this gas economically. This paper examines two methods of producing gas from hydrate deposits by the injection of hot water or steam, and also examines the feasibility of hydraulic fracturing and pressure reduction as a hydrate gas production technique. A hydraulic fracturing technique suitable for hydrate reservoirs and a system for coring hydrate reservoirs are also described.

  17. Strategies for gas production from oceanic Class 3 hydrateaccumulations

    SciTech Connect

    Moridis, George J.; Reagan, Matthew T.

    2007-05-01

    Gas hydrates are solid crystalline compounds in which gasmolecules are lodged within the lattices of ice crystals. Vast amounts ofCH4 are trapped in gas hydrates, and a significant effort has recentlybegun to evaluate hydrate deposits as a potential energy source. Class 3hydrate deposits are characterized by an isolated Hydrate-Bearing Layer(HBL) that is not in contact with any hydrate-free zone of mobile fluids.The base of the HBL in Class 3 deposits may occur within or at the edgeof the zone of thermodynamic hydrate stability.In this numerical study oflong-term gas production from typical representatives of unfracturedClass 3 deposits, we determine that simple thermal stimulation appears tobe a slow and inefficient production method. Electrical heating and warmwater injection result in very low production rates (4 and 12 MSCFD,respectively) that are orders of magnitude lower than generallyacceptable standards of commercial viability of gas production fromoceanic reservoirs. However, production from depressurization-baseddissociation based on a constant well pressure appears to be a promisingapproach even in deposits characterized by high hydrate saturations. Thisapproach allows the production of very large volumes ofhydrate-originating gas at high rates (>15 MMSCFD, with a long-termaverage of about 8.1 MMSCFD for the reference case) for long times usingconventional technology. Gas production from hydrates is accompanied by asignificant production of water. However, unlike conventional gasreservoirs, the water production rate declines with time. The lowsalinity of the produced water may require care in its disposal. Becauseof the overwhelming advantage of depressurization-based methods, thesensitivity analysis was not extendedto thermal stimulation methods. Thesimulation results indicate that depressurization-induced gas productionfrom oceanic Class 3 deposits increases (and the corresponding waterto-gas ratio decreases) with increasing hydrate temperature (whichdefines the hydrate stability), increasing intrinsic permeability of theHBL, and decreasing hydrate saturation although depletion of the hydratemay complicate the picture in the latter case.

  18. Process for production desulfurized of synthesis gas

    DOEpatents

    Wolfenbarger, James K. (Torrance, CA); Najjar, Mitri S. (Wappingers Falls, NY)

    1993-01-01

    A process for the partial oxidation of a sulfur- and silicate-containing carbonaceous fuel to produce a synthesis gas with reduced sulfur content which comprises partially oxidizing said fuel at a temperature in the range of 1900.degree.-2600.degree. F. in the presence of a temperature moderator, an oxygen-containing gas and a sulfur capture additive which comprises a calcium-containing compound portion, a sodium-containing compound portion, and a fluoride-containing compound portion to produce a synthesis gas comprising H.sub.2 and CO with a reduced sulfur content and a molten slag which comprises (1) a sulfur-containing sodium-calcium-fluoride silicate phase; and (2) a sodium-calcium sulfide phase.

  19. 40 CFR Table W - 1A of Subpart W-Default Whole Gas Emission Factors for Onshore Petroleum and Natural Gas Production

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...Whole Gas Emission Factors for Onshore Petroleum and Natural Gas Production W Table... MANDATORY GREENHOUSE GAS REPORTING Petroleum and Natural Gas Systems Definitions...Whole Gas Emission Factors for Onshore Petroleum and Natural Gas Production...

  20. Outer Continental Shelf Oil and Gas Leasing/Production Program

    SciTech Connect

    Not Available

    1988-01-01

    This annual report on the Outer Continental Shelf (OCS) Oil and Gas Leasing and Production program summarizes receipts and expenditures, and includes information on OCS safety violations as reported by the US Coast Guard. 3 figs., 12 tabs.

  1. Tempest gas turbine extends EGT product line

    SciTech Connect

    Chellini, R.

    1995-07-01

    With the introduction of the 7.8 MW (mechanical output) Tempest gas turbine, ECT has extended the company`s line of its small industrial turbines. The new Tempest machine, featuring a 7.5 MW electric output and a 33% thermal efficiency, ranks above the company`s single-shaft Typhoon gas turbine, rated 3.2 and 4.9 MW, and the 6.3 MW Tornado gas turbine. All three machines are well-suited for use in combined heat and power (CHP) plants, as demonstrated by the fact that close to 50% of the 150 Typhoon units sold are for CHP applications. This experience has induced EGT, of Lincoln, England, to announce the introduction of the new gas turbine prior to completion of the testing program. The present single-shaft machine is expected to be used mainly for industrial trial cogeneration. This market segment, covering the needs of paper mills, hospitals, chemical plants, ceramic industry, etc., is a typical local market. Cogeneration plants are engineered according to local needs and have to be assisted by local organizations. For this reason, to efficiently cover the world market, EGT has selected a number of associates that will receive from Lincoln completely engineered machine packages and will engineer the cogeneration system according to custom requirements. These partners will also assist the customer and dispose locally of the spares required for maintenance operations.

  2. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  3. Yrast Excitations around Doubly Magic {sup {bold 132}}Sn from Fission Product {bold {gamma}}-Ray Studies

    SciTech Connect

    Zhang, C.T.; Bhattacharyya, P.; Daly, P.J.; Broda, R.; Grabowski, Z.W.; Nisius, D.; Ahmad, I.; Ishii, T.; Carpenter, M.P.; Morss, L.R.; Phillips, W.R.; Durell, J.L.; Leddy, M.J.; Smith, A.G.; Urban, W.; Varley, B.J.; Schulz, N.; Lubkiewicz, E.; Bentaleb, M.; Blomqvist, J.

    1996-10-01

    Prompt {gamma}-ray cascades in neutron-rich nuclei around doubly magic {sup 132}Sn have been studied at Eurogam II using a {sup 248}Cm fission source. Yrast states to above 5.5 MeV in the two- and three-proton {ital N}=82 isotones {sup 134}Te and {sup 135}I are reported. They are interpreted in terms of valence proton and particle-hole core excitations with the help of shell model calculations employing empirical nucleon-nucleon interactions from both {sup 132}Sn and {sup 208}Pb regions. A serious inconsistency in the accepted masses of {ital N}=82 isotones near {sup 132}Sn is discovered but not resolved. {copyright} {ital 1996 The American Physical Society.}

  4. 21 CFR 886.5918 - Rigid gas permeable contact lens care products.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ...2010-04-01 false Rigid gas permeable contact lens care products. 886.5918... § 886.5918 Rigid gas permeable contact lens care products. (a) Identification. A rigid gas permeable contact lens care product is a device...

  5. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...I determine the royalty due on gas production? 1202.550 Section...DEPARTMENT OF THE INTERIOR Natural Resources Revenue ROYALTIES Gas Production From Indian Leases...I determine the royalty due on gas production? If you...

  6. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ...2010-07-01 false How do I measure natural gas production on my eligible lease... § 260.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  7. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...2012-07-01 false How do I measure natural gas production on my eligible lease... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  8. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...2014-07-01 false How do I measure natural gas production on my eligible lease... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  9. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...2013-07-01 false How do I measure natural gas production on my eligible lease... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  10. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...2011-07-01 false How do I measure natural gas production on my eligible lease... § 260.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible...

  11. Fission Yield Measurements by Inductively Coupled Plasma Mass-Spectrometry

    SciTech Connect

    Irina Glagolenko; Bruce Hilton; Jeffrey Giglio; Daniel Cummings; Karl Grimm; Richard McKnight

    2009-11-01

    Correct prediction of the fission products inventory in irradiated nuclear fuels is essential for accurate estimation of fuel burnup, establishing proper requirements for spent fuel transportation and storage, materials accountability and nuclear forensics. Such prediction is impossible without accurate knowledge of neutron induced fission yields. Unfortunately, the accuracy of the fission yields reported in the ENDF/B-VII.0 library is not uniform across all of the data and much of the improvement is desired for certain isotopes and fission products. We discuss our measurements of cumulative fission yields in nuclear fuels irradiated in thermal and fast reactor spectra using Inductively Coupled Plasma Mass Spectrometry.

  12. Challenges, uncertainties and issues facing gas production from gas hydrate deposits

    SciTech Connect

    Moridis, G.J.; Collett, T.S.; Pooladi-Darvish, M.; Hancock, S.; Santamarina, C.; Boswell, R.; Kneafsey, T.; Rutqvist, J.; Kowalsky, M.; Reagan, M.T.; Sloan, E.D.; Sum, A.K.; Koh, C.

    2010-11-01

    The current paper complements the Moridis et al. (2009) review of the status of the effort toward commercial gas production from hydrates. We aim to describe the concept of the gas hydrate petroleum system, to discuss advances, requirement and suggested practices in gas hydrate (GH) prospecting and GH deposit characterization, and to review the associated technical, economic and environmental challenges and uncertainties, including: the accurate assessment of producible fractions of the GH resource, the development of methodologies for identifying suitable production targets, the sampling of hydrate-bearing sediments and sample analysis, the analysis and interpretation of geophysical surveys of GH reservoirs, well testing methods and interpretation of the results, geomechanical and reservoir/well stability concerns, well design, operation and installation, field operations and extending production beyond sand-dominated GH reservoirs, monitoring production and geomechanical stability, laboratory investigations, fundamental knowledge of hydrate behavior, the economics of commercial gas production from hydrates, and the associated environmental concerns.

  13. Thorium-uranium fission radiography

    NASA Technical Reports Server (NTRS)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  14. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  15. Elemental Fluorine-18 Gas: Enhanced Production and Availability

    SciTech Connect

    VanBrocklin, Henry F.

    2011-12-01

    The overall objective of this project was to develop an efficient, reproducible and reliable process for the preparation of fluorine-18 labeled fluorine gas ([¹?F]F?) from readily available cyclotron-produced [¹?F]fluoride ion. The two step process entailed the production of [¹?F]fluoromethane with subsequent conversion to [¹?F]F? by electric discharge of [¹?F]fluoromethane in the presence of carrier nonradioactive F? gas. The specific goals of this project were i) to optimize the preparation of [¹?F]fluoromethane from [¹?F]fluoride ion; ii) to develop a prototype automated system for the production of [¹?F]F? from [¹?F]fluoride ion and iii) develop a compact user friendly automated system for the preparation of [¹?F]F? with initial synthesis of fluorine-18 labeled radiotracers. Over the last decade there has been an increased interest in the production of "non-standard" positron-emitting isotopes for the preparation of new radiotracers for a variety of applications including medical imaging and therapy. The increased availability of these isotopes from small biomedical cyclotrons has prompted their use in labeling radiotracers. In much the same way the production of [¹?F]F? gas has been known for several decades. However, access to [¹?F]F? gas has been limited to those laboratories with the means (e.g. F? targetry for the cyclotron) and the project-based need to work with [¹?F]F? gas. Relatively few laboratories, compared to those that produce [¹?F]fluoride ion on a daily basis, possess the capability to produce and use [¹?F]F? gas. A simplified, reliable system employing [¹?F]fluoride ion from cyclotron targetry systems that are already in place coupled with on-demand production of the [¹?F]F? gas would greatly enhance its availability. This would improve the availability of [¹?F]F? gas and promote further work with a valuable precursor. The major goals of the project were accomplished over the funding period. The preparation of ¹?F]fluoromethane has been automated with reproducible yields greater than 90% conversion from [¹?F]fluoride ion. A trap and release system was established for the [¹?F]fluoride ion concentration and direct elution of the [¹?F]fluoride ion into the reaction vial with the appropriate base and precursor in DMSO. Other solvents were also investigated. The production time for [¹?F]fluoromethane is less than 10 minutes. An automated system for the [¹?F]F? gas production from the [18F]fluoromethane has been developed. The unit coupled to the [¹?F]fluoromethane system permits the on demand production of [¹?F]F? gas. In less than 30 minutes, mCi quantities of [¹?F]F? gas were produced. Several variables for the [¹?F]F? gas production were investigated and a set of parameters for reproducible operation were determined. These parameters included discharge chamber size, carrier gas (He, Ne, Ar), discharge time, discharge current, mass of F? gas added to the chamber. FDOPA and EF5 were used to test the reactivity of the [¹?F]F? gas. Both products were produced in low to modest yield. The ready availability of [¹?F]F? gas has potential impact to advance both DOE mission-driven initiatives and nuclear medicine initiatives through other federally funded agencies such as NIH and DoD. New reactions involving the use of [¹?F]F? gas will lead to direct labeling of new radiotracers and intermediates as well as new fluorine-18 labeled synthons that may be further reacted with other reagents to provide useful fluorine-18 labeled compounds. New tracers to understand and follow plant and microbial metabolism as well as new tracers for nuclear medicine applications, that have been either difficult to obtain or never produced due to the limited availability of [¹?F]F? gas, may be prepared using the techniques developed .

  16. Fission gas release behaviour of a 103 GWd/tHM fuel disc during a 1200 °C annealing test

    NASA Astrophysics Data System (ADS)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.; Tverberg, T.

    2014-03-01

    Within the Nuclear Fuel Industry Research (NFIR) program, several fuel variants, in the form of thin circular discs, were irradiated in the Halden Boiling Water Reactor (HBWR) to a range of burn-ups ˜100 GWd/tHM. The design of the assembly was similar to that used in other HBWR programs: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature gradients within the fuel discs. One such rod contained standard grain UO2 discs (3D grain size = 18 ?m) reaching a burn-up of 103 GWd/tHM. After the irradiation, the gas release upon rod puncturing was measured to be 2.9%.

  17. Preliminary report on the commercial viability of gas production from natural gas hydrates

    USGS Publications Warehouse

    Walsh, M.R.; Hancock, S.H.; Wilson, S.J.; Patil, S.L.; Moridis, G.J.; Boswell, R.; Collett, T.S.; Koh, C.A.; Sloan, E.D.

    2009-01-01

    Economic studies on simulated gas hydrate reservoirs have been compiled to estimate the price of natural gas that may lead to economically viable production from the most promising gas hydrate accumulations. As a first estimate, $CDN2005 12/Mscf is the lowest gas price that would allow economically viable production from gas hydrates in the absence of associated free gas, while an underlying gas deposit will reduce the viability price estimate to $CDN2005 7.50/Mscf. Results from a recent analysis of the simulated production of natural gas from marine hydrate deposits are also considered in this report; on an IROR basis, it is $US2008 3.50-4.00/Mscf more expensive to produce marine hydrates than conventional marine gas assuming the existence of sufficiently large marine hydrate accumulations. While these prices represent the best available estimates, the economic evaluation of a specific project is highly dependent on the producibility of the target zone, the amount of gas in place, the associated geologic and depositional environment, existing pipeline infrastructure, and local tariffs and taxes. ?? 2009 Elsevier B.V.

  18. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    SciTech Connect

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  19. Report on possible routes to breakdown products of mustard gas

    SciTech Connect

    Luman, F.M.

    1983-10-18

    This paper suggests possible routes to the formation of decontamination and breakdown products of the chemical agent Mustard Gas (HD). The terminal decontamination products, CaSO4 and CO2, are harmless to the environment. Oxathiane is formed by hydrolysis and dehydration reactions. Dithiane is formed with the application of heat in a low oxygen or nitrogen environment. (Author).

  20. NEUTRAL PRODUCTS FROM GAS PHASE REARRANGEMENTS OF SIMPLE

    E-print Network

    Morton, Thomas Hellman

    NEUTRAL PRODUCTS FROM GAS PHASE REARRANGEMENTS OF SIMPLE CARBOCATIONS Thomas Hellman Morton OUTLINE . . . . . . . . . . . . . . . . . . . . . . 217 C. Distinguishing Transposition from Randomization . . . . . . . . . 221 D. Formation of Ion­Neutral Elsevier Science B.V. All rights reserved. #12;ABSTRACT Analyzing the neutral products from ionic reactions

  1. Spontaneous and induced emission of XeCl* excimer molecules under pumping of Xe – CCl{sub 4} and Ar – Xe – CCl{sub 4} gas mixtures with a low CCl{sub 4} content by fast electrons and uranium fission fragments

    SciTech Connect

    Mis'kevich, A I; Guo, J; Dyuzhov, Yu A

    2013-11-30

    The spontaneous and induced emission of XeCl* excimer molecules upon excitation of Xe – CCl{sub 4} and Ar – Xe – CCl{sub 4} gas mixtures with a low CCl{sub 4} content by high-energy charged particles [a pulsed high-energy electron beam and products of neutron nuclear reaction {sup 235}U(n, f)] has been experimentally studied. The electron energy was 150 keV, and the pump current pulse duration and amplitude were 5 ns and 5 A, respectively. The energy of fission fragments did not exceed 100 MeV, the duration of the neutron pump pulse was 200 ?s, and the specific power contribution to the gas was about 300 W cm{sup -3}. Electron beam pumping in a cell 4 cm long with a cavity having an output mirror transmittance of 2.7% gives rise to lasing on the B ? X transition in the XeCl* molecule (? = 308 nm) with a gain ? = 0.0085 cm{sup -1} and fluorescence efficiency ? ? 10%. Pumping by fission fragments in a 250-cm-long cell with a cavity formed by a highly reflecting mirror and a quartz window implements amplified spontaneous emission (ASE) with an output power of 40 – 50 kW sr{sup -1} and a base ASE pulse duration of ?200 ms. .

  2. US production of natural gas from tight reservoirs

    SciTech Connect

    Not Available

    1993-10-18

    For the purposes of this report, tight gas reservoirs are defined as those that meet the Federal Energy Regulatory Commission`s (FERC) definition of tight. They are generally characterized by an average reservoir rock permeability to gas of 0.1 millidarcy or less and, absent artificial stimulation of production, by production rates that do not exceed 5 barrels of oil per day and certain specified daily volumes of gas which increase with the depth of the reservoir. All of the statistics presented in this report pertain to wells that have been classified, from 1978 through 1991, as tight according to the FERC; i.e., they are ``legally tight`` reservoirs. Additional production from ``geologically tight`` reservoirs that have not been classified tight according to the FERC rules has been excluded. This category includes all producing wells drilled into legally designated tight gas reservoirs prior to 1978 and all producing wells drilled into physically tight gas reservoirs that have not been designated legally tight. Therefore, all gas production referenced herein is eligible for the Section 29 tax credit. Although the qualification period for the credit expired at the end of 1992, wells that were spudded (began to be drilled) between 1978 and May 1988, and from November 5, 1990, through year end 1992, are eligible for the tax credit for a subsequent period of 10 years. This report updates the EIA`s tight gas production information through 1991 and considers further the history and effect on tight gas production of the Federal Government`s regulatory and tax policy actions. It also provides some high points of the geologic background needed to understand the nature and location of low-permeability reservoirs.

  3. Fission dynamics within time-dependent Hartree-Fock: boost-induced fission

    E-print Network

    Goddard, P M; Rios, A

    2015-01-01

    Background: Nuclear fission is a complex large-amplitude collective decay mode in heavy nuclei. Microscopic density functional studies of fission have previously concentrated on adiabatic approaches based on constrained static calculations ignoring dynamical excitations of the fissioning nucleus, and the daughter products. Purpose: To explore the ability of dynamic mean-field methods to describe induced fission processes, using quadrupole boosts in the nuclide $^{240}$Pu as an example. Methods: Quadrupole constrained Hartree-Fock calculations are used to create a potential energy surface. An isomeric state and a state beyond the second barrier peak are excited by means of instantaneous as well as temporally extended gauge boosts with quadrupole shapes. The subsequent deexcitation is studied in a time-dependent Hartree-Fock simulation, with emphasis on fissioned final states. The corresponding fission fragment mass numbers are studied. Results: In general, the energy deposited by the quadrupole boost is quickl...

  4. Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model

    E-print Network

    Diecker, Jane T

    2005-01-01

    The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

  5. Compact fission counter for DANCE

    SciTech Connect

    Wu, C Y; Chyzh, A; Kwan, E; Henderson, R; Gostic, J; Carter, D; Bredeweg, T; Couture, A; Jandel, M; Ullmann, J

    2010-11-06

    The Detector for Advanced Neutron Capture Experiments (DANCE) consists of 160 BF{sub 2} crystals with equal solid-angle coverage. DANCE is a 4{pi} {gamma}-ray calorimeter and designed to study the neutron-capture reactions on small quantities of radioactive and rare stable nuclei. These reactions are important for the radiochemistry applications and modeling the element production in stars. The recognition of capture event is made by the summed {gamma}-ray energy which is equivalent of the reaction Q-value and unique for a given capture reaction. For a selective group of actinides, where the neutron-induced fission reaction competes favorably with the neutron capture reaction, additional signature is needed to distinguish between fission and capture {gamma} rays for the DANCE measurement. This can be accomplished by introducing a detector system to tag fission fragments and thus establish a unique signature for the fission event. Once this system is implemented, one has the opportunity to study not only the capture but also fission reactions. A parallel-plate avalanche counter (PPAC) has many advantages for the detection of heavy charged particles such as fission fragments. These include fast timing, resistance to radiation damage, and tolerance of high counting rate. A PPAC also can be tuned to be insensitive to {alpha} particles, which is important for experiments with {alpha}-emitting actinides. Therefore, a PPAC is an ideal detector for experiments requiring a fast and clean trigger for fission. A PPAC with an ingenious design was fabricated in 2006 by integrating amplifiers into the target assembly. However, this counter was proved to be unsuitable for this application because of issues related to the stability of amplifiers and the ability to separate fission fragments from {alpha}'s. Therefore, a new design is needed. A LLNL proposal to develop a new PPAC for DANCE was funded by NA22 in FY09. The design goal is to minimize the mass for the proposed counter and still be able to maintain a stable operation under extreme radioactivity and the ability to separate fission fragments from {alpha}'s. In the following sections, the description is given for the design and performance of this new compact PPAC, for studying the neutron-induced reactions on actinides using DANCE at LANL.

  6. Production of Substitute Natural Gas from Coal

    SciTech Connect

    Andrew Lucero

    2009-01-31

    The goal of this research program was to develop and demonstrate a novel gasification technology to produce substitute natural gas (SNG) from coal. The technology relies on a continuous sequential processing method that differs substantially from the historic methanation or hydro-gasification processing technologies. The thermo-chemistry relies on all the same reactions, but the processing sequences are different. The proposed concept is appropriate for western sub-bituminous coals, which tend to be composed of about half fixed carbon and about half volatile matter (dry ash-free basis). In the most general terms the process requires four steps (1) separating the fixed carbon from the volatile matter (pyrolysis); (2) converting the volatile fraction into syngas (reforming); (3) reacting the syngas with heated carbon to make methane-rich fuel gas (methanation and hydro-gasification); and (4) generating process heat by combusting residual char (combustion). A key feature of this technology is that no oxygen plant is needed for char combustion.

  7. Event-by-event fission simulation code, generates complete fission events

    Energy Science and Technology Software Center (ESTSC)

    2013-04-01

    FREYA is a computer code that generates complete fission events. The output includes the energy and momentum of these final state particles: fission products, prompt neutrons and prompt photons. The version of FREYA that is to be released is a module for MCNP6.

  8. Effects of gas composition in headspace and bicarbonate concentrations in media on gas and methane production, degradability, and rumen fermentation using in vitro gas production techniques.

    PubMed

    Patra, Amlan Kumar; Yu, Zhongtang

    2013-07-01

    Headspace gas composition and bicarbonate concentrations in media can affect methane production and other characteristics of rumen fermentation in in vitro gas production systems, but these 2 important factors have not been evaluated systematically. In this study, these 2 factors were investigated with respect to gas and methane production, in vitro digestibility of feed substrate, and volatile fatty acid (VFA) profile using in vitro gas production techniques. Three headspace gas compositions (N2+ CO2+ H2 in the ratio of 90:5:5, CO2, and N2) with 2 substrate types (alfalfa hay only, and alfalfa hay and a concentrate mixture in a 50:50 ratio) in a 3×2 factorial design (experiment 1) and 3 headspace compositions (N2, N2 + CO2 in a 50:50 ratio, and CO2) with 3 bicarbonate concentrations (80, 100, and 120 mM) in a 3×3 factorial design (experiment 2) were evaluated. In experiment 1, total gas production (TGP) and net gas production (NGP) was the lowest for CO2, followed by N2, and then the gas mixture. Methane concentration in headspace gas after fermentation was greater for CO2 than for N2 and the gas mixture, whereas total methane production (TMP) and net methane production (NMP) were the greatest for CO2, followed by the gas mixture, and then N2. Headspace composition did not affect in vitro digestibility or the VFA profile, except molar percentages of propionate, which were greater for CO2 and N2 than for the gas mixture. Methane concentration in headspace gas, TGP, and NGP were affected by the interaction of headspace gas composition and substrate type. In experiment 2, increasing concentrations of CO2 in the headspace decreased TGP and NGP quadratically, but increased the concentrations of methane, NMP, and in vitro fiber digestibility linearly, and TMP quadratically. Fiber digestibility, TGP, and NGP increased linearly with increasing bicarbonate concentrations in the medium. Concentrations of methane and NMP were unaffected by bicarbonate concentration, but TMP tended to increase due to increasing bicarbonate concentration. Although total VFA concentration and molar percentage of butyrate were unchanged, the molar percentage of acetate, and acetate-to-propionate ratio decreased, whereas the molar percentage of propionate increased quadratically with increasing bicarbonate concentration. This study demonstrated for the first time that headspace composition, especially CO2 content, and bicarbonate concentration in media could significantly influence gas and methane production, and rumen fermentation in gas production techniques. PMID:23684023

  9. Natural gas productive capacity for the lower 48 states 1985 through 1997

    SciTech Connect

    1996-12-01

    This publication presents information on wellhead productive capacity and a projection of gas production requirements. A history of natural gas production and productive capacity at the wellhead, along with a projection of the same, is illustrated.

  10. Evaluation of the gas production economics of the gas hydrate cyclic thermal injection model

    SciTech Connect

    Kuuskraa, V.A.; Hammersheimb, E.; Sawyer, W.

    1985-05-01

    The objective of the work performed under this directive is to assess whether gas hydrates could potentially be technically and economically recoverable. The technical potential and economics of recovering gas from a representative hydrate reservoir will be established using the cyclic thermal injection model, HYDMOD, appropriately modified for this effort, integrated with economics model for gas production on the North Slope of Alaska, and in the deep offshore Atlantic. The results from this effort are presented in this document. In Section 1, the engineering cost and financial analysis model used in performing the economic analysis of gas production from hydrates -- the Hydrates Gas Economics Model (HGEM) -- is described. Section 2 contains a users guide for HGEM. In Section 3, a preliminary economic assessment of the gas production economics of the gas hydrate cyclic thermal injection model is presented. Section 4 contains a summary critique of existing hydrate gas recovery models. Finally, Section 5 summarizes the model modification made to HYDMOD, the cyclic thermal injection model for hydrate gas recovery, in order to perform this analysis.

  11. On-Board Hydrogen Gas Production System For Stirling Engines

    SciTech Connect

    Johansson, Lennart N.

    2004-06-29

    A hydrogen production system for use in connection with Stirling engines. The production system generates hydrogen working gas and periodically supplies it to the Stirling engine as its working fluid in instances where loss of such working fluid occurs through usage through operation of the associated Stirling engine. The hydrogen gas may be generated by various techniques including electrolysis and stored by various means including the use of a metal hydride absorbing material. By controlling the temperature of the absorbing material, the stored hydrogen gas may be provided to the Stirling engine as needed. A hydrogen production system for use in connection with Stirling engines. The production system generates hydrogen working gas and periodically supplies it to the Stirling engine as its working fluid in instances where loss of such working fluid occurs through usage through operation of the associated Stirling engine. The hydrogen gas may be generated by various techniques including electrolysis and stored by various means including the use of a metal hydride absorbing material. By controlling the temperature of the absorbing material, the stored hydrogen gas may be provided to the Stirling engine as needed.

  12. Gas Production from Hydrate-Bearing Sediments - Emergent Phenomena -

    SciTech Connect

    Jung, J.W.; Jang, J.W.; Tsouris, Costas; Phelps, Tommy Joe; Rawn, Claudia J; Santamarina, Carlos

    2012-01-01

    Even a small fraction of fine particles can have a significant effect on gas production from hydrate-bearing sediments and sediment stability. Experiments were conducted to investigate the role of fine particles on gas production using a soil chamber that allows for the application of an effective stress to the sediment. This chamber was instrumented to monitor shear-wave velocity, temperature, pressure, and volume change during CO{sub 2} hydrate formation and gas production. The instrumented chamber was placed inside the Oak Ridge National Laboratory Seafloor Process Simulator (SPS), which was used to control the fluid pressure and temperature. Experiments were conducted with different sediment types and pressure-temperature histories. Fines migrated within the sediment in the direction of fluid flow. A vuggy structure formed in the sand; these small cavities or vuggs were precursors to the development of gas-driven fractures during depressurization under a constant effective stress boundary condition. We define the critical fines fraction as the clay-to-sand mass ratio when clays fill the pore space in the sand. Fines migration, clogging, vugs, and gas-driven fracture formation developed even when the fines content was significantly lower than the critical fines fraction. These results show the importance of fines in gas production from hydrate-bearing sediments, even when the fines content is relatively low.

  13. Water Resources and Natural Gas Production from the Marcellus Shale

    USGS Publications Warehouse

    Soeder, Daniel J.; Kappel, William M.

    2009-01-01

    The Marcellus Shale is a sedimentary rock formation deposited over 350 million years ago in a shallow inland sea located in the eastern United States where the present-day Appalachian Mountains now stand (de Witt and others, 1993). This shale contains significant quantities of natural gas. New developments in drilling technology, along with higher wellhead prices, have made the Marcellus Shale an important natural gas resource. The Marcellus Shale extends from southern New York across Pennsylvania, and into western Maryland, West Virginia, and eastern Ohio (fig. 1). The production of commercial quantities of gas from this shale requires large volumes of water to drill and hydraulically fracture the rock. This water must be recovered from the well and disposed of before the gas can flow. Concerns about the availability of water supplies needed for gas production, and questions about wastewater disposal have been raised by water-resource agencies and citizens throughout the Marcellus Shale gas development region. This Fact Sheet explains the basics of Marcellus Shale gas production, with the intent of helping the reader better understand the framework of the water-resource questions and concerns.

  14. Theoretical Description of the Fission Process

    SciTech Connect

    Witold Nazarewicz

    2009-10-25

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation’s nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic waste and be proliferation-resistant, is a goal for the advanced nuclear fuel cycles program. While in the past the design, construction, and operation of reactors were supported through empirical trials, this new phase in nuclear energy production is expected to heavily rely on advanced modeling and simulation capabilities.

  15. Cascade heat recovery with coproduct gas production

    DOEpatents

    Brown, W.R.; Cassano, A.A.; Dunbobbin, B.R.; Rao, P.; Erickson, D.C.

    1986-10-14

    A process for the integration of a chemical absorption separation of oxygen and nitrogen from air with a combustion process is set forth wherein excess temperature availability from the combustion process is more effectively utilized to desorb oxygen product from the absorbent and then the sensible heat and absorption reaction heat is further utilized to produce a high temperature process stream. The oxygen may be utilized to enrich the combustion process wherein the high temperature heat for desorption is conducted in a heat exchange preferably performed with a pressure differential of less than 10 atmospheres which provides considerable flexibility in the heat exchange. 4 figs.

  16. Cascade heat recovery with coproduct gas production

    DOEpatents

    Brown, William R. (Zionsville, PA); Cassano, Anthony A. (Allentown, PA); Dunbobbin, Brian R. (Allentown, PA); Rao, Pradip (Allentown, PA); Erickson, Donald C. (Annapolis, MD)

    1986-01-01

    A process for the integration of a chemical absorption separation of oxygen and nitrogen from air with a combustion process is set forth wherein excess temperature availability from the combustion process is more effectively utilized to desorb oxygen product from the absorbent and then the sensible heat and absorption reaction heat is further utilized to produce a high temperature process stream. The oxygen may be utilized to enrich the combustion process wherein the high temperature heat for desorption is conducted in a heat exchange preferably performed with a pressure differential of less than 10 atmospheres which provides considerable flexibility in the heat exchange.

  17. A1. SHALE GAS PRODUCTION GROWTH IN THE UNITED STATES..............................1 A2. VARIABILITY IN SHALE WELL PRODUCTION PERFORMANCE ............................1

    E-print Network

    1 APPENDIX1 Contents A1. SHALE GAS PRODUCTION GROWTH IN THE UNITED STATES FOR FLOWBACK GAS CAPTURE IN SHALE PLAYS..9 A5. REFERENCES...................................................................................................................13 A1. SHALE GAS PRODUCTION GROWTH IN THE UNITED STATES Natural gas production in the United States

  18. Forecasting Gas Production in Organic Shale with the Combined Numerical Simulation of Gas Diffusion in Kerogen, Langmuir Desorption from

    E-print Network

    Torres-Verdín, Carlos

    SPE 159250 Forecasting Gas Production in Organic Shale with the Combined Numerical Simulation algorithm to forecast gas production in organic shale that simultaneously takes into account gas diffusion-than-expected permeability in shale-gas formations, while Langmuir desorption maintains pore pressure. Simulations confirm

  19. Engineering analysis of biomass gasifier product gas cleaning technology

    SciTech Connect

    Baker, E.G.; Brown, M.D.; Moore, R.H.; Mudge, L.K.; Elliott, D.C.

    1986-08-01

    For biomass gasification to make a significant contribution to the energy picture in the next decade, emphasis must be placed on the generation of clean, pollutant-free gas products. This reports attempts to quantify levels of particulated, tars, oils, and various other pollutants generated by biomass gasifiers of all types. End uses for biomass gases and appropriate gas cleaning technologies are examined. Complete systems analysis is used to predit the performance of various gasifier/gas cleanup/end use combinations. Further research needs are identified. 128 refs., 20 figs., 19 tabs.

  20. Fifty years with nuclear fission. Volume 2

    SciTech Connect

    Behrens, J.W.; Carlson, A.D.

    1989-12-31

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, ``Fifty years with nuclear fission,`` in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately.

  1. Phase Transition and Fragment Production in the Lattice Gas Model

    NASA Astrophysics Data System (ADS)

    Gulminelli, Francesca; Chomaz, Philippe

    The critical behavior of fragment production is studied within a Lattice Gas Model in the canonical ensemble. Finite size effects on the liquid-gas phase transition are analyzed by a direct calculation of the partition function, and it is shown that phase coexistence and phase transition are relevant concepts even for systems of a few tens of particles. Critical exponents are extracted from the behavior of the fragment production yield as a function of temperature by means of a finite size scaling. The result is that in a finite system well defined critical signals can be found at supercritical (Kertész line) as well as subcritical densities inside the coexistence zone.

  2. Gas production from oceanic Class 2 hydrate accumulations

    SciTech Connect

    Moridis, G.J.; Reagan, M.T.

    2007-02-01

    Gas hydrates are solid crystalline compounds in which gasmolecules are lodged within the lattices of ice crystals. The vastamounts of hydrocarbon gases that are trapped in hydrate deposits in thepermafrost and in deep ocean sediments may constitute a promising energysource. Class 2 hydrate deposits are characterized by a Hydrate-BearingLayer (HBL) that is underlain by a saturated zone of mobile water. Inthis study we investigated three methods of gas production via verticalwell designs. A long perforated interval (covering the hydrate layer andextending into the underlying water zone) yields the highest gasproduction rates (up to 20 MMSCFD), but is not recommended for long-termproduction because of severe flow blockage caused by secondary hydrateand ice. A short perforated interval entirely within the water zoneallows long-term production, but only at rates of 4.5 7 MMSCFD. A newwell design involving localized heating appears to be the most promising,alleviating possible blockage by secondary hydrate and/or ice near thewellbore) and delivering sustainably large, long-term rates (10-15MMSCFD).The production strategy involves a cyclical process. During eachcycle, gas production continuously increases, while the correspondingwater production continuously decreases. Each cycle is concluded by acavitation event (marked by a precipitous pressure drop at the well),brought about by the inability of thesystem to satisfy the constant massproduction rate QM imposed at the well. This is caused by the increasinggas contribution to the production stream, and/or flow inhibition causedby secondary hydrate and/or ice. In the latter case, short-term thermalstimulation removes the blockage. The results show that gas productionincreases (and the corresponding water-to-gas ratio RWGC decreases) withan increasing(a) QM, (b) hydrate temperature (which defines its stabilityfor a given pressure), and (c) intrinsic permeability. Lower initialhydrate saturations lead initially to higher gas production and a lowerRWGC, but the effect is later reversed as the hydrate is depleted. Thedisposal of the large amounts of produced water does not appear to pose asignificant environmental problem. Production from Class 2 hydrates ischaracterized by (a) the need for confining boundaries, (b) thecontinuously improving RWGC over time (opposite to conventional gasreservoirs), and (c) the development of a free gas zone at the top of thehydrate layer (necessitating the existence of a gas cap forproduction).

  3. Kinetics study on biomass pyrolysis for fuel gas production.

    PubMed

    Chen, Guan-Yi; Fang, Meng-Xiang; Andries, J; Luo, Zhong-Yang; Spliethoff, H; Cen, Ke-Fa

    2003-01-01

    Kinetic knowledge is of great importance in achieving good control of the pyrolysis and gasification process and optimising system design. An overall kinetic pyrolysis scheme is therefore addressed here. The kinetic modelling incorporates the following basic steps: the degradation of the virgin biomass materials into primary products (tar, gas and semi-char), the decomposition of primary tar into secondary products and the continuous interaction between primary gas and char. The last step is disregarded completely by models in the literature. Analysis and comparison of predicted results from different kinetic schemes and experimental data on our fixed bed pyrolyser yielded very positive evidence to support our kinetic scheme. PMID:12861621

  4. Depressurization and electrical heating of hydrate sediment for gas production

    NASA Astrophysics Data System (ADS)

    Minagawa, H.; Ito, T.; Kimura, S.; Kaneko, H.; Noda, S.; Narita, H.

    2014-12-01

    In-situ dissociation of natural gas hydrate is necessary for commercial recovery of natural gas from natural gas hydrate sediment. Thermal stimulation is an effective dissociation method, along with depressurization. In this study, we examined the efficiency of electrical heating of the hydrate core for gas production. In order to evaluate efficiency of electrical heating with depressurization, we investigated following subject. (1) electrical heating of Xe gas hydrate sediment, as a conventional simulation of methane hydrate sediment, (2) electrical heating of methane hydrate sediment, which was compared with Xe gas hydrate experiment, and (3) electrical heating of hydrate bearing sediment with fine sandy layer which was simulated faults with large displacement shear around hydrate sediment. These experiments revealed that depressurization and additional electrode heating of hydrate sediment saturated with electrolyte solution was confirmed to enable higher efficient and effective gas production from sedimentwith less electric power. This study is financially supported by METI and Research Consortium for Methane Hydrate Resources in Japan (the MH21 Research Consortium).

  5. Natural gas production problems : solutions, methodologies, and modeling.

    SciTech Connect

    Rautman, Christopher Arthur; Herrin, James M.; Cooper, Scott Patrick; Basinski, Paul M.; Olsson, William Arthur; Arnold, Bill Walter; Broadhead, Ronald F.; Knight, Connie D.; Keefe, Russell G.; McKinney, Curt; Holm, Gus; Holland, John F.; Larson, Rich; Engler, Thomas W.; Lorenz, John Clay

    2004-10-01

    Natural gas is a clean fuel that will be the most important domestic energy resource for the first half the 21st centtuy. Ensuring a stable supply is essential for our national energy security. The research we have undertaken will maximize the extractable volume of gas while minimizing the environmental impact of surface disturbances associated with drilling and production. This report describes a methodology for comprehensive evaluation and modeling of the total gas system within a basin focusing on problematic horizontal fluid flow variability. This has been accomplished through extensive use of geophysical, core (rock sample) and outcrop data to interpret and predict directional flow and production trends. Side benefits include reduced environmental impact of drilling due to reduced number of required wells for resource extraction. These results have been accomplished through a cooperative and integrated systems approach involving industry, government, academia and a multi-organizational team within Sandia National Laboratories. Industry has provided essential in-kind support to this project in the forms of extensive core data, production data, maps, seismic data, production analyses, engineering studies, plus equipment and staff for obtaining geophysical data. This approach provides innovative ideas and technologies to bring new resources to market and to reduce the overall environmental impact of drilling. More importantly, the products of this research are not be location specific but can be extended to other areas of gas production throughout the Rocky Mountain area. Thus this project is designed to solve problems associated with natural gas production at developing sites, or at old sites under redevelopment.

  6. Organic Sulfur Gas Production in Sulfidic Caves

    NASA Astrophysics Data System (ADS)

    Stern, L. A.; Engel, A. S.; Bennett, P. C.

    2001-12-01

    Lower Kane Cave, Big Horn Basin, WY, permits access to an environment where anaerobic sulfide-rich groundwater meets the aerobic vadose zone. At this interface microorganisms thrive on diverse metabolic pathways including autotrophic sulfur oxidation, sulfate reduction, and aerobic heterotrophy. Springs introduce groundwater rich in H2S to the cave where it both degasses into the cave atmosphere and is used by chemautotrophic sulfur oxidizing bacteria in the cave spring and stream habitat. The cave atmosphere in the immediate vicinity of the springs has elevated levels of CO2, H2S and methane, mirroring the higher concentration of H2S and methane in the spring water. The high CO2 concentrations are attenuated toward the two main sources of fresh air, the cave entrance and breathing holes at the rear of the cave. Conventional toxic gas monitors permit estimations of H2S concentrations, but they have severe cross sensitivity with other reduced sulfur gases, and thus are inadequate for characterization of sulfur cave gases. However employment of a field-based GC revealed elevated concentrations of carbonyl sulfide in cave atmosphere. Cultures of microorganisms collected from the cave optimized for enriching fermenters and autotrophic and heterophic sulfate reducing bacteria each produced carbonyl sulfide suggesting a biogenic in origin of the COS in addition to H2S. Enrichment cultures also produced methanethiol (methyl mercaptan) and an additional as yet undetermined volatile organic sulfur compound. In culture, the organo-sulfur compounds were less abundant than H2S, whereas in the cave atmosphere the organo-sulfur compounds were the dominant sulfur gases. Thus, these organo-sulfur gases may prove to be important sources of both reduced sulfur and organic carbon to microorganisms living on the cave wall in a subaerial habitat. Moreover groundwater has not yet been recognized as a source of sulfur gases to the atmosphere, but with the abundance of sulfidic groundwater, this environment may prove to be important to the global sulfur cycle and its influence of the global radiation budget.

  7. Synthesis gas production by mixed conducting membranes with integrated conversion into liquid products

    DOEpatents

    Nataraj, Shankar (Allentown, PA); Russek, Steven Lee (Allentown, PA); Dyer, Paul Nigel (Allentown, PA)

    2000-01-01

    Natural gas or other methane-containing feed gas is converted to a C.sub.5 -C.sub.19 hydrocarbon liquid in an integrated system comprising an oxygenative synthesis gas generator, a non-oxygenative synthesis gas generator, and a hydrocarbon synthesis process such as the Fischer-Tropsch process. The oxygenative synthesis gas generator is a mixed conducting membrane reactor system and the non-oxygenative synthesis gas generator is preferably a heat exchange reformer wherein heat is provided by hot synthesis gas product from the mixed conducting membrane reactor system. Offgas and water from the Fischer-Tropsch process can be recycled to the synthesis gas generation system individually or in combination.

  8. Fifty years with nuclear fission. Volume 1

    SciTech Connect

    Behrens, J.W.; Carlson, A.D.

    1989-12-31

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, ``Fifty Years with Nuclear Fission,`` in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately.

  9. Complete and Incomplete Fusion Competition in 11B-INDUCED Fission Reactions on 197Au at the Intermediate Energy

    NASA Astrophysics Data System (ADS)

    Demekhina, N. A.; Karapetyan, G. S.; Balabekyan, A. R.

    2015-06-01

    Above Coulomb barrier cross sections of fission fragment production were measured in reactions of 11B with 197Au target. Induced-activity method was used for measurement the fission decay channel of the composite nuclei. Systematic of the fission fragment charge and mass distributions was used for fission cross section calculation. Fission fraction of the composite nuclei decay was compared with PACE-4 mode calculations. Estimated suppression for fission fraction followed the complete fusion have been obtained 35%.

  10. Electrical systems for oil and gas production facilities

    SciTech Connect

    Bishop, D.N.

    1988-01-01

    The design of electrical systems associated with oil and gas production facilities requires knowledge of specialized subjects. This book discusses the important and unique features of the major portions of electrical systems for these facilities. Also, certain safety considerations specifically applicable to production facilities are addressed in some detail, particularly installations in locations that might be exposed to ignitable concentrations of flammable gases or vapors.

  11. Entropy Production and Thermal Conductivity of A Dilute Gas

    E-print Network

    Yong-Jun Zhang

    2011-02-16

    It is known that the thermal conductivity of a dilute gas can be derived by using kinetic theory. We present here a new derivation by starting with two known entropy production principles: the steepest entropy ascent (SEA) principle and the maximum entropy production (MEP) principle. A remarkable feature of the new derivation is that it does not require the specification of the existence of the temperature gradient. The known result is reproduced in a similar form.

  12. BUILDING MATERIALS MADE FROM FLUE GAS DESULFURIZATION BY-PRODUCTS

    SciTech Connect

    Michael W. Grutzeck; Maria DiCola; Paul Brenner

    2006-03-30

    Flue gas desulphurization (FGD) materials are produced in abundant quantities by coal burning utilities. Due to environmental restrains, flue gases must be ''cleaned'' prior to release to the atmosphere. They are two general methods to ''scrub'' flue gas: wet and dry. The choice of scrubbing material is often defined by the type of coal being burned, i.e. its composition. Scrubbing is traditionally carried out using a slurry of calcium containing material (slaked lime or calcium carbonate) that is made to contact exiting flue gas as either a spay injected into the gas or in a bubble tower. The calcium combined with the SO{sub 2} in the gas to form insoluble precipitates. Some plants have been using dry injection of these same materials or their own Class C fly ash to scrub. In either case the end product contains primarily hannebachite (CaSO{sub 3} {center_dot} 1/2H{sub 2}O) with smaller amounts of gypsum (CaSO{sub 4} {center_dot} 2H{sub 2}O). These materials have little commercial use. Experiments were carried out that were meant to explore the feasibility of using blends of hannebachite and fly ash mixed with concentrated sodium hydroxide to make masonry products. The results suggest that some of these mixtures could be used in place of conventional Portland cement based products such as retaining wall bricks and pavers.

  13. Production of bio-synthetic natural gas in Canada.

    PubMed

    Hacatoglu, Kevork; McLellan, P James; Layzell, David B

    2010-03-15

    Large-scale production of renewable synthetic natural gas from biomass (bioSNG) in Canada was assessed for its ability to mitigate energy security and climate change risks. The land area within 100 km of Canada's network of natural gas pipelines was estimated to be capable of producing 67-210 Mt of dry lignocellulosic biomass per year with minimal adverse impacts on food and fiber production. Biomass gasification and subsequent methanation and upgrading were estimated to yield 16,000-61,000 Mm(3) of pipeline-quality gas (equivalent to 16-63% of Canada's current gas use). Life-cycle greenhouse gas emissions of bioSNG-based electricity were calculated to be only 8.2-10% of the emissions from coal-fired power. Although predicted production costs ($17-21 GJ(-1)) were much higher than current energy prices, a value for low-carbon energy would narrow the price differential. A bioSNG sector could infuse Canada's rural economy with $41-130 billion of investments and create 410,000-1,300,000 jobs while developing a nation-wide low-carbon energy system. PMID:20175525

  14. Role of dynamical effects in the formation of T-Odd asymmetries for products of polarized-neutron-induced ternary fission of nuclei

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

    Basic dynamical effects that accompany the cold-polarized-neutron-induced binary and ternary fission of actinide nuclei and which determine the properties of T -odd asymmetries in angular distributions of various prescission and evaporated light third particles emitted in true and delayed ternary fission are analyzed on the basis of quantum-mechanical fission theory. It is emphasized that effects associated with the conservation of axial symmetry of the fissioning system under study at all stages of its evolution from the formation of neutron resonance states of the fissile compound nucleus to the separation of its fission fragments, including the appearance of zero wriggling vibrations of the cold compound nucleus in the vicinity of its scission point, are of particular importance, the influence of quantum collective rotation of the polarized fissile system on the asymmetry of the angular distribution of both fission fragments and third particles being taken into account. It is shown that the difference in the behavior of the coefficients characterizing the T -odd asymmetries under analysis for the target nuclei being studied can be explained, upon taking into account the interference between the fission amplitudes for the neutron resonance states of fissile compound nuclei, by the difference in the contributions of even and odd components of the amplitudes of angular distributions of third particles to the coefficients in question.

  15. Investigation of gas production and entrapment in granular iron medium

    NASA Astrophysics Data System (ADS)

    Kamolpornwijit, W.; Liang, L.

    2006-01-01

    A method for measuring gas entrapment in granular iron (Fe 0) was developed and used to estimate the impact of gas production on porosity loss during the treatment of a high NO 3- groundwater (up to ˜10 mM). Over the 400-d study period the trapped gas in laboratory columns was small, with a maximum measured at 1.3% pore volume. Low levels of dissolved H 2(g) were measured (up to 0.07 ± 0.02 M). Free moving gas bubbles were not observed. Thus, porosity loss, which was determined by tracer tests to be 25-30%, is not accounted for by residual gas trapped in the iron. The removal of aqueous species (i.e., NO 3-, Ca, and carbonate alkalinity) indicates that mineral precipitation contributed more significantly to porosity loss than did the trapped gases. Using the stoichiometric reactions between Fe 0 and NO 3-, an average corrosion rate of 1.7 mmol kg - 1 d - 1 was derived for the test granular iron. This rate is 10 times greater than Fe 0 oxidation by H 2O alone, based on H 2 gas production. NO 3- ion rather than H 2O was the major oxidant in the groundwater in the absence of molecular O 2. The N-mass balance [e.g., N 2(g) and NH 4+ and NO 3-] suggests that abiotic reduction of NO 3- dominated at the start of Fe 0 treatment, whereas N 2 production became more important once the microbial activity began. These laboratory results closely predict N 2 gas production in a separated large column experiment that was operated for ˜2 yr in the field, where a maximum of ˜600 ml d - 1 gas volumes was detected, of which 99.5% (v/v) was N 2. We conclude that NO 3- suppressed the production of H 2(g) by competing with water for Fe 0 oxidation, especially at the beginning of water treatment when Fe 0 is highly reactive. Depends on the groundwater composition, gas venting may be necessary in maintaining PRB performance in the field.

  16. Trace gas flux from container production of woody landscape plants

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The agriculture industry is a large source of greenhouse gas (GHG) emissions which are widely believed to be causing increased global temperatures. Reduction of these emissions has been heavily researched, with most of the work focusing on row crop and animal production sectors. Little attention has...

  17. Fifty years of nuclear fission: Nuclear data and measurements series

    SciTech Connect

    Lynn, J.E.

    1989-06-01

    This report is the written version of a colloquium first presented at Argonne National Laboratory in January 1989. The paper begins with an historical preamble about the events leading to the discovery of nuclear fission. This leads naturally to an account of early results and understanding of the fission phenomena. Some of the key concepts in the development of fission theory are then discussed. The main theme of this discussion is the topography of the fission barrier, in which the interplay of the liquid-drop model and nucleon shell effects lead to a wide range of fascinating phenomena encompassing metastable isomers, intermediate-structure effects in fission cross-sections, and large changes in fission product properties. It is shown how study of these changing effects and theoretical calculations of the potential energy of the deformed nucleus have led to broad qualitative understanding of the nature of the fission process. 54 refs., 35 figs.

  18. Deepwater production drives design of new Gulf gas plant

    SciTech Connect

    Nielsen, R.A.; Petty, L.; Elliot, D.; Chen, R.

    1998-03-16

    Exploration and production success in deepwater, eastern Gulf of Mexico has created the need for additional gas-transmission and processing infrastructure. The Destin pipeline and the Pascagoula gas-processing plant are being built to serve this need. The Destin pipeline originates at a junction platform at Main Pass 260 and, after coming ashore near Pascagoula, Miss., will connect with five interstate gas-transmission pipelines, by-passing gas-transportation bottlenecks in Louisiana and Alabama. The Pascagoula plant will be built near the point the pipeline comes ashore and immediately before the first compressor station. The paper discusses handling condensate, design goals, achieving objectives, low life-cycle cost, and project schedule.

  19. Environmental Compliance for Oil and Gas Exploration and Production

    SciTech Connect

    Hansen, Christine

    1999-10-26

    The Appalachian/Illinois Basin Directors is a group devoted to increasing communication among the state oil and gas regulatory agencies within the Appalachian and Illinois Basin producing region. The group is comprised of representatives from the oil and gas regulatory agencies from states in the basin (Attachment A). The directors met to discuss regulatory issues common to the area, organize workshops and seminars to meet the training needs of agencies dealing with the uniqueness of their producing region and perform other business pertinent to this area of oil and gas producing states. The emphasis of the coordinated work was a wide range of topics related to environmental compliance for natural gas and oil exploration and production.

  20. Alaska North Slope regional gas hydrate production modeling forecasts

    USGS Publications Warehouse

    Wilson, S.J.; Hunter, R.B.; Collett, T.S.; Hancock, S.; Boswell, R.; Anderson, B.J.

    2011-01-01

    A series of gas hydrate development scenarios were created to assess the range of outcomes predicted for the possible development of the "Eileen" gas hydrate accumulation, North Slope, Alaska. Production forecasts for the "reference case" were built using the 2002 Mallik production tests, mechanistic simulation, and geologic studies conducted by the US Geological Survey. Three additional scenarios were considered: A "downside-scenario" which fails to identify viable production, an "upside-scenario" describes results that are better than expected. To capture the full range of possible outcomes and balance the downside case, an "extreme upside scenario" assumes each well is exceptionally productive.Starting with a representative type-well simulation forecasts, field development timing is applied and the sum of individual well forecasts creating the field-wide production forecast. This technique is commonly used to schedule large-scale resource plays where drilling schedules are complex and production forecasts must account for many changing parameters. The complementary forecasts of rig count, capital investment, and cash flow can be used in a pre-appraisal assessment of potential commercial viability.Since no significant gas sales are currently possible on the North Slope of Alaska, typical parameters were used to create downside, reference, and upside case forecasts that predict from 0 to 71??BM3 (2.5??tcf) of gas may be produced in 20 years and nearly 283??BM3 (10??tcf) ultimate recovery after 100 years.Outlining a range of possible outcomes enables decision makers to visualize the pace and milestones that will be required to evaluate gas hydrate resource development in the Eileen accumulation. Critical values of peak production rate, time to meaningful production volumes, and investments required to rule out a downside case are provided. Upside cases identify potential if both depressurization and thermal stimulation yield positive results. An "extreme upside" case captures the full potential of unconstrained development with widely spaced wells. The results of this study indicate that recoverable gas hydrate resources may exist in the Eileen accumulation and that it represents a good opportunity for continued research. ?? 2010 Elsevier Ltd.

  1. Analytical Modeling of Shale Hydraulic Fracturing and Gas Production

    NASA Astrophysics Data System (ADS)

    Xu, W.

    2012-12-01

    Shale gas is abundant all over the world. Due to its extremely low permeability, extensive stimulation of a shale reservoir is always required for its economic production. Hydraulic fracturing has been the primary method of shale reservoir stimulation. Consequently the design and optimization of a hydraulic fracturing treatment plays a vital role insuring job success and economic production. Due to the many variables involved and the lack of a simple yet robust tool based on fundamental physics, horizontal well placement and fracturing job designs have to certain degree been a guessing game built on previous trial and error experience. This paper presents a method for hydraulic fracturing design and optimization in these environments. The growth of a complex hydraulic fracture network (HFN) during a fracturing job is equivalently represented by a wiremesh fracturing model (WFM) constructed on the basis of fracture mechanics and mass balance. The model also simulates proppant transport and placement during HFN growth. Results of WFM simulations can then be used as the input into a wiremesh production model (WPM) constructed based on WFM. WPM represents gas flow through the wiremesh HFN by an elliptic flow and the flow of gas in shale matrix by a novel analytical solution accounting for contributions from both free and adsorbed gases stored in the pore space. WPM simulation is validated by testing against numerical simulations using a commercially available reservoir production simulator. Due to the analytical nature of WFM and WPM, both hydraulic fracturing and gas production simulations run very fast on a regular personal computer and are suitable for hydraulic fracturing job design and optimization. A case study is presented to demonstrate how a non-optimized hydraulic fracturing job might have been optimized using WFM and WPM simulations.Fig. 1. Ellipsoidal representation of (a) stimulated reservoir and (b) hydraulic fracture network created by hydraulic fracturing treatment. Fig. 2. Gas flow represented by (a) elliptical flow through fracture network and (b) linear flow within reservoir matrix.

  2. Potential Operating Orbits for Fission Electric Propulsion Systems Driven by the SAFE-400

    NASA Technical Reports Server (NTRS)

    Houts, Mike; Kos, Larry; Poston, David; Rodgers, Stephen L. (Technical Monitor)

    2002-01-01

    Safety must be ensured during all phases of space fission system design, development, fabrication, launch, operation, and shutdown. One potential space fission system application is fission electric propulsion (FEP), in which fission energy is converted into electricity and used to power high efficiency (Isp greater than 3000s) electric thrusters. For these types of systems it is important to determine which operational scenarios ensure safety while allowing maximum mission performance and flexibility. Space fission systems are essentially nonradioactive at launch, prior to extended operation at high power. Once high power operation begins, system radiological inventory steadily increases as fission products build up. For a given fission product isotope, the maximum radiological inventory is typically achieved once the system has operated for a length of time equivalent to several half-lives. After that time, the isotope decays at the same rate it is produced, and no further inventory builds in. For an FEP mission beginning in Earth orbit, altitude and orbital lifetime increase as the propulsion system operates. Two simultaneous effects of fission propulsion system operation are thus (1) increasing fission product inventory and (2) increasing orbital lifetime. Phrased differently, as fission products build up, more time is required for the fission products to naturally convert back into non-radioactive isotopes. Simultaneously, as fission products build up, orbital lifetime increases, providing more time for the fission products to naturally convert back into non-radioactive isotopes. Operational constraints required to ensure safety can thus be quantified.

  3. Trash-to-Gas: Converting Space Trash into Useful Products

    NASA Technical Reports Server (NTRS)

    Caraccio, Anne J.; Hintze, Paul E.

    2013-01-01

    NASA's Logistical Reduction and Repurposing (LRR) project is a collaborative effort in which NASA is determined to reduce total logistical mass through reduction, reuse and recycling of various wastes and components of long duration space missions and habitats. LRR is focusing on four distinct advanced areas of study: Advanced Clothing System, Logistics-to-Living, Heat Melt Compactor and Trash to Supply Gas (TtSG). The objective of TtSG is to develop technologies that convert material waste, human waste and food waste into high-value products. High-value products include life support oxygen and water, rocket fuels, raw material production feedstocks, and other energy sources. There are multiple pathways for converting waste to products involving single or multi-step processes. This paper discusses thermal oxidation methods of converting waste to methane. Different wastes, including food, food packaging, Maximum Absorbent Garments (MAGs), human waste simulants, and cotton washcloths have been evaluated in a thermal degradation reactor under conditions promoting pyrolysis, gasification or incineration. The goal was to evaluate the degradation processes at varying temperatures and ramp cycles and to maximize production of desirable products and minimize high molecular weight hydrocarbon (tar) production. Catalytic cracking was also evaluated to minimize tar production. The quantities of CO2, CO, CH4, and H2O were measured under the different thermal degradation conditions. The conversion efficiencies of these products were used to determine the best methods for producing desired products.

  4. Fission products in National Atmospheric Deposition Program—Wet deposition samples prior to and following the Fukushima Dai-Ichi Nuclear Power Plant incident, March 8?April 5, 2011

    USGS Publications Warehouse

    Wetherbee, Gregory A.; Debey, Timothy M.; Nilles, Mark A.; Lehmann, Christopher M.B.; Gay, David A.

    2012-01-01

    Radioactive isotopes I-131, Cs-134, or Cs-137, products of uranium fission, were measured at approximately 20 percent of 167 sampled National Atmospheric Deposition Program monitoring sites in North America (primarily in the contiguous United States and Alaska) after the Fukushima Dai-Ichi Nuclear Power Plant incident on March 12, 2011. Samples from the National Atmospheric Deposition Program were analyzed for the period of March 8-April 5, 2011. Calculated 1- or 2-week radionuclide deposition fluxes at 35 sites from Alaska to Vermont ranged from 0.47 to 5,100 Becquerels per square meter during the sampling period of March 15-April 5, 2011. No fission-product isotopes were measured in National Atmospheric Deposition Program samples obtained during March 8-15, 2011, prior to the arrival of contaminated air in North America.

  5. NOBLE GAS PRODUCTION FROM MERCURY SPALLATION AT SNS

    SciTech Connect

    DeVore, Joe R; Lu, Wei; Schwahn, Scott O

    2013-01-01

    Calculations for predicting the distribution of the products of spallation reactions between high energy protons and target materials are well developed and are used for design and operational applications in many projects both within DOE and in other arenas. These calculations are based on theory and limited experimental data that verifies rates of production of some spallation products exist. At the Spallation Neutron Source, a helium stream from the mercury target flows through a system to remove radioactivity from this mercury target offgas. The operation of this system offers a window through which the production of noble gases from mercury spallation by protons may be observed. This paper describes studies designed to measure the production rates of twelve noble gas isotopes within the Spallation Neutron Source mercury target.

  6. Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast Breeder Reactor Blankets 

    E-print Network

    Osborn, Jeremy

    2014-08-13

    A nuclear terrorist attack is one of the most serious threats to the national security of the United States, and in the wake of an attack, attribution of responsibility will be of the utmost importance. Plutonium, a by-product in spent nuclear...

  7. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 2 2010-07-01 2010-07-01 false How do I measure natural gas production on my... do I measure natural gas production on my eligible lease? You must measure natural gas production on... natural gas, measured according to part 250, subpart L of this title, equals one barrel of oil...

  8. The Neutron Induced Fission Fragment Tracking Experiment: Hardware Overview

    NASA Astrophysics Data System (ADS)

    Duke, Dana; Niffte Collaboration

    2011-04-01

    The goal of the Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) is to measure fission cross sections with unprecedented precision using time projection chamber (TPC) technology. The NIFFTE TPC hardware and supporting systems are discussed, including neutron source, gas system, and data acquisition. The current status of hardware implementation for the experiment will be presented.

  9. Gas plant economic optimization is more than meeting product specification

    SciTech Connect

    Berkowitz, P.N.; Colwell, L.W.; Gamez, J.P.

    1996-12-31

    Gas plants require a higher level of process control to optimize the process to maximize operating profits. Automation alone does not achieve this objective whereas, on-line dynamic optimization of the control variables based on product pricing, the cost to process the gas and the contracts for gas and liquids is solvable by new control techniques. Daily operations are affected by a paradigm shift in the method of control for the facility. This newly developed and site proven technique has demonstrated how to improve benefits when net processing margins are positive and minimize operating cost when liquids margins are negative. Because ethane recovery versus its rejection is not a binary decision, a better means to operate can be shown to benefit the gas plant operator. Each specification has a cost to meet it or a penalty to exceed it. However, if allowed, exceeding specification may prove beneficial to the net profitability of the operations. With the decision being made on-line every few minutes, the results are more dramatic than previously understood. Gas Research Institute and Continental Controls, Inc. have installed more than 10 such systems in US gas processing plants. Project payout from the use of the MVC{reg_sign} technology has on average been less than six months. Processing savings have ranged from $.0075 to $.024 per Mcf. The authors paper last year showed where the benefits can be derived. This year the results of those facilities are shared along with the methodology to achieve them.

  10. A double ionization chamber for fission fragment detection

    NASA Astrophysics Data System (ADS)

    Chepigin, V. I.; Stepantsov, S. V.; Nagy, S.; Ter-Akopian, G. M.; Voronin, A. S.

    1984-03-01

    In the present paper some problems associated with studies of the spontaneous fission of transfermium elements are considered, and advantages of ionization chambers over other detectors for fission fragment detection are discussed. A double ionization chamber for detecting fission fragments has been built, and the energy and mass calibrations have been performed using the thermal neutron-induced fission of 235U. The use of a gas jet, in combination with a double ionization chamber as a detector, in experiments with heavy ions is discussed. Model experiments have been carried out.

  11. Monitoring system for a liquid-cooled nuclear fission reactor

    DOEpatents

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  12. Halogens in oil and gas production-associated wastewater.

    NASA Astrophysics Data System (ADS)

    Harkness, J.; Warner, N. R.; Dwyer, G. S.; Mitch, W.; Vengosh, A.

    2014-12-01

    Elevated chloride and bromide in oil and gas wastewaters that are released to the environment are one of the major environmental risks in areas impacted by shale gas development [Olmstead et al.,2013]. In addition to direct contamination of streams, the potential for formation of highly toxic disinfection by-products (DBPs) in drinking water in utilities located downstream from disposal sites poses a serious risk to human health. Here we report on the occurrence of iodide in oil and gas wastewater. We conducted systematic measurements of chloride, bromide, and iodide in (1) produced waters from conventional oil and gas wells from the Appalachian Basin; (2) hydraulic fracturing flowback fluids from unconventional Marcellus and Fayetteville shale gas, (3) effluents from a shale gas spill site in West Virginia; (4) effluents of oil and gas wastewater disposed to surface water from three brine treatment facilities in western Pennsylvania; and (5) surface waters downstream from the brine treatment facilities. Iodide concentration was measured by isotope dilution-inductively coupled plasma-mass spectrometry, which allowed for a more accurate measurement of iodide in a salt-rich matrix. Iodide in both conventional and unconventional oil and gas produced and flowback waters varied from 1 mg/L to 55 mg/L, with no systematic enrichment in hydraulic fracturing fluids. The similarity in iodide content between the unconventional Marcellus flowback waters and the conventional Appalachian produced waters clearly indicate that the hydraulic fracturing process does not induce additional iodide and the iodide content is related to natural variations in the host formations. Our data show that effluents from the brine treatment facilities have elevated iodide (mean = 20.9±1 mg/L) compared to local surface waters (0.03± 0.1 mg/L). These results indicate that iodide, in addition to chloride and bromide in wastewater from oil and gas production, poses an additional risk to downstream surface water quality and drinking water utilities given the potential of formation of iodate-DBPs in drinking water. Olmstead, S.M. et al. (2013). Shale gas development impacts on surface water quality in Pennsylvania, PNAS, 110, 4962-4967.

  13. A multiple parallel-plate avalanche counter for fission-fragment detection

    NASA Astrophysics Data System (ADS)

    Wu, C. Y.; Henderson, R. A.; Haight, R. C.; Lee, H. Y.; Taddeucci, T. N.; Bucher, B.; Chyzh, A.; Devlin, M.; Fotiades, N.; Kwan, E.; O'Donnell, J. M.; Perdue, B. A.; Ullmann, J. L.

    2015-09-01

    A new low-mass multiple gas-filled parallel-plate avalanche counter for the fission-fragment detection has been developed to mark the fission occurrence in measurements of the prompt fission neutron energy spectrum as a function of incident neutron energy. It was used successfully for the neutron-induced fission of 235U and 239Pu with a total mass near 100 mg each and the spontaneous fission of 252Cf. Both the incident neutron energy and the prompt fission neutron energy are measured by using the time-of-flight method. The design and performance of this avalanche counter are described.

  14. Biomodal spontaneous fission

    SciTech Connect

    Hulet, E.K. )

    1989-09-26

    Investigations of mass and kinetic-energy distributions from spontaneous fission have been extended in recent years to an isotope of element 104 and, for half-lives, to an isotope of element 108. The results have been surprising in that spontaneous fission half-lives have turned out to be much longer than expected and mass and kinetic- energy distributions were found to abruptly shift away from those of the lighter actinides, showing two modes of fission. These new developments have caused a re-evaluation of our understanding of the fission process, bringing an even deeper appreciation of the role played by nuclear shell effects upon spontaneous fission properties. 16 refs., 10 figs.

  15. Electroplating fission-recoil barriers onto LEU-metal foils for {sup 99}Mo-production targets

    SciTech Connect

    Smaga, J.A.; Sedlet, J.; Conner, C.; Liberatore, M.W.; Walker, D.E.; Wygmans, D.G.; Vandegrift, G.F.

    1997-10-01

    Electroplating experiments on uranium foil have been conducted in order to develop low-enriched uranium composite targets suitable for the production of {sup 99}Mo. Preparation of the foil surface prior to plating was found to play a key role in the quality of the resultant coating. A surface preparation procedure was developed that produces both zinc and nickel coatings with the desired level of coating adherence and coverage. Modifications of the existing plating processes now need investigation to improve to uniformity of the plating thickness, especially at the foil perimeter.

  16. Radiolytic gas production in the alpha particle degradation of plastics

    SciTech Connect

    Reed, D.T.; Hoh, J.; Emery, J. ); Hobbs, D. )

    1992-01-01

    Net gas generation due to alpha particle irradiation of polyethylene and polyvinyl chloride was investigated. Experiments were performed in an air environment at 30, 60, and 100{degree}C. The predominant radiolytic degradation products of polyethylene were hydrogen and carbon dioxide with a wide variety of trace organic species noted. Irradiation of polyvinyl chloride resulted in the formation of HCl in addition to the products observed for polyethylene. For both plastic materials, a strong enhancement of net yields was noted at 100{degree}C.

  17. Radiolytic gas production in the alpha particle degradation of plastics

    SciTech Connect

    Reed, D.T.; Hoh, J.; Emery, J.; Hobbs, D.

    1992-05-01

    Net gas generation due to alpha particle irradiation of polyethylene and polyvinyl chloride was investigated. Experiments were performed in an air environment at 30, 60, and 100{degree}C. The predominant radiolytic degradation products of polyethylene were hydrogen and carbon dioxide with a wide variety of trace organic species noted. Irradiation of polyvinyl chloride resulted in the formation of HCl in addition to the products observed for polyethylene. For both plastic materials, a strong enhancement of net yields was noted at 100{degree}C.

  18. Challenges, uncertainties, and issues facing gas production from gas-hydrate deposits

    USGS Publications Warehouse

    Moridis, G.J.; Collett, T.S.; Pooladi-Darvish, M.; Hancock, S.; Santamarina, C.; Boswel, R.; Kneafsey, T.; Rutqvist, J.; Kowalsky, M.B.; Reagan, M.T.; Sloan, E.D.; Sum, A.K.; Koh, C.A.

    2011-01-01

    The current paper complements the Moridis et al. (2009) review of the status of the effort toward commercial gas production from hydrates. We aim to describe the concept of the gas-hydrate (GH) petroleum system; to discuss advances, requirements, and suggested practices in GH prospecting and GH deposit characterization; and to review the associated technical, economic, and environmental challenges and uncertainties, which include the following: accurate assessment of producible fractions of the GH resource; development of methods for identifying suitable production targets; sampling of hydrate-bearing sediments (HBS) and sample analysis; analysis and interpretation of geophysical surveys of GH reservoirs; well-testing methods; interpretation of well-testing results; geomechanical and reservoir/well stability concerns; well design, operation, and installation; field operations and extending production beyond sand-dominated GH reservoirs; monitoring production and geomechanical stability; laboratory investigations; fundamental knowledge of hydrate behavior; the economics of commercial gas production from hydrates; and associated environmental concerns. ?? 2011 Society of Petroleum Engineers.

  19. Production of light oil by injection of hot inert gas

    NASA Astrophysics Data System (ADS)

    Ruidas, Bidhan C.; Ganguly, Somenath

    2015-07-01

    Hot inert gas, when injected into an oil reservoir is capable of generating a vaporization-condensation drive and as a consequence, a preferential movement of the lighter components to the production well. This form of displacement is an important unit mechanism in hot flue-gas injection, or in thermal recovery from a watered-out oil reservoir. This article presents the movement of heat front vis-à-vis the changes in the saturation profile, and the gas-phase composition. The plateau in the temperature profile due to the exchange of latent heat, and the formation of water bank at the downstream are elaborated. The broadening of the vaporization-condensation zone with continued progression is discussed. The effect of inert gas temperature on the cumulative production of oil is reviewed. The results provide insight to the vaporization-condensation drive as a stand-alone mechanism. The paper underscores the relative importance of this mechanism, when operated in tandem with other processes in improved oil recovery and CO2 sequestration.

  20. Dissolved gas exsolution to enhance gas production and transport during bench-scale electrical resistance heating

    NASA Astrophysics Data System (ADS)

    Hegele, P. R.; Mumford, K. G.

    2015-05-01

    Condensation of volatile organic compounds in colder zones can be detrimental to the performance of an in situ thermal treatment application for the remediation of chlorinated solvent source zones. A novel method to increase gas production and limit convective heat loss in more permeable, potentially colder, zones involves the injection and liberation of dissolved gas from solution during heating. Bench-scale electrical resistance heating experiments were performed with a dissolved carbon dioxide and sodium chloride solution to investigate exsolved gas saturations and transport regimes at elevated, but sub-boiling, temperatures. At sub-boiling temperatures, maximum exsolved gas saturations of Sg = 0.12 were attained, and could be sustained when the carbon dioxide solution was injected during heating rather than emplaced prior to heating. This gas saturation was estimated to decrease groundwater relative permeability to krw = 0.64. Discontinuous gas transport was observed above saturations of Sg = 0.07, demonstrating the potential of exsolved CO2 to bridge vertical gas transport through colder zones.

  1. Organic Substances from Unconventional Oil and Gas Production in Shale

    NASA Astrophysics Data System (ADS)

    Orem, W. H.; Varonka, M.; Crosby, L.; Schell, T.; Bates, A.; Engle, M.

    2014-12-01

    Unconventional oil and gas (UOG) production has emerged as an important element in the US and world energy mix. Technological innovations in the oil and gas industry, especially horizontal drilling and hydraulic fracturing, allow for the enhanced release of oil and natural gas from shale compared to conventional oil and gas production. This has made commercial exploitation possible on a large scale. Although UOG is enormously successful, there is surprisingly little known about the effects of this technology on the targeted shale formation and on environmental impacts of oil and gas production at the surface. We examined water samples from both conventional and UOG shale wells to determine the composition, source and fate of organic substances present. Extraction of hydrocarbon from shale plays involves the creation and expansion of fractures through the hydraulic fracturing process. This process involves the injection of large volumes of a water-sand mix treated with organic and inorganic chemicals to assist the process and prop open the fractures created. Formation water from a well in the New Albany Shale that was not hydraulically fractured (no injected chemicals) had total organic carbon (TOC) levels that averaged 8 mg/L, and organic substances that included: long-chain fatty acids, alkanes, polycyclic aromatic hydrocarbons, heterocyclic compounds, alkyl benzenes, and alkyl phenols. In contrast, water from UOG production in the Marcellus Shale had TOC levels as high as 5,500 mg/L, and contained a range of organic chemicals including, solvents, biocides, scale inhibitors, and other organic chemicals at thousands of ?g/L for individual compounds. These chemicals and TOC decreased rapidly over the first 20 days of water recovery as injected fluids were recovered, but residual organic compounds (some naturally-occurring) remained up to 250 days after the start of water recovery (TOC 10-30 mg/L). Results show how hydraulic fracturing changes the organic composition of shale formation water, and that some injected organic substances are retained on the shale and slowly released. Thus, appropriate safe disposal of produced water is needed long into production. Changes in organic substances in formation water may impact microbial communities. Current work is focused on UOG production in the Permian Basin, Texas.

  2. Thermal Flammable Gas Production from Bulk Vitrification Feed

    SciTech Connect

    Scheele, Randall D.; McNamara, Bruce K.; Bagaasen, Larry M.

    2008-05-21

    The baseline bulk-vitrification (BV) process (also known as in-container vitrification ICV™) includes a mixer/dryer to convert liquid low-activity waste (LAW) into a dried, blended feed for vitrification. Feed preparation includes blending LAW with glass-forming minerals (GFMs) and cellulose and drying the mixture to a suitable dryness, consistency, and particle size for transport to the ICVTM container. The cellulose is to be added to the BV feed at a rate sufficient to destroy 75% of the nitrogen present as nitrate or nitrite. Concern exists that flammable gases may be produced during drying operations at levels that could pose a risk. The drying process is conducted under vacuum in the temperature range of 60 to 80°C. These flammable gases could be produced either through thermal decomposition of cellulose or waste organics or as a by-product of the reaction of cellulose and/or waste organics with nitrate or the postulated small amount of nitrite present in the waste. To help address the concern about flammable gas production during drying, the Pacific Northwest National Laboratory (PNNL) performed studies to identify the gases produced at dryer temperatures and at possible process upset conditions. Studies used a thermogravimetric analyzer (TGA) up to 525°C and isothermal testing up to 120°C to determine flammable gas production resulting from the cellulose and organic constituents in bulk vitrification feed. This report provides the results of those studies to determine the effects of cellulose and waste organics on flammable gas evolution

  3. Gas Sensor Evaluations in Polymer Combustion Product Atmospheres

    NASA Technical Reports Server (NTRS)

    Delgado, Rafael H.; Davis, Dennis D.; Beeson, Harold D.

    1999-01-01

    Toxic gases produced by the combustion or thermo-oxidative degradation of materials such as wire insulation, foam, plastics, or electronic circuit boards in space shuttle or space station crew cabins may pose a significant hazard to the flight crew. Toxic gas sensors are routinely evaluated in pure gas standard mixtures, but the possible interferences from polymer combustion products are not routinely evaluated. The NASA White Sands Test Facility (WSTF) has developed a test system that provides atmospheres containing predetermined quantities of target gases combined with the coincidental combustion products of common spacecraft materials. The target gases are quantitated in real time by infrared (IR) spectroscopy and verified by grab samples. The sensor responses are recorded in real time and are compared to the IR and validation analyses. Target gases such as carbon monoxide, hydrogen cyanide, hydrogen chloride, and hydrogen fluoride can be generated by the combustion of poly(vinyl chloride), polyimide-fluoropolymer wire insulation, polyurethane foam, or electronic circuit board materials. The kinetics and product identifications for the combustion of the various materials were determined by thermogravimetric-IR spectroscopic studies. These data were then scaled to provide the required levels of target gases in the sensor evaluation system. Multisensor toxic gas monitors from two manufacturers were evaluated using this system. In general, the sensor responses satisfactorily tracked the real-time concentrations of toxic gases in a dynamic mixture. Interferences from a number of organic combustion products including acetaldehyde and bisphenol-A were minimal. Hydrogen bromide in the products of circuit board combustion registered as hydrogen chloride. The use of actual polymer combustion atmospheres for the evaluation of sensors can provide additional confidence in the reliability of the sensor response.

  4. Production of bioplastics and hydrogen gas by photosynthetic microorganisms

    NASA Astrophysics Data System (ADS)

    Yasuo, Asada; Masato, Miyake; Jun, Miyake

    1998-03-01

    Our efforts have been aimed at the technological basis of photosynthetic-microbial production of materials and an energy carrier. We report here accumulation of poly-(3-hydroxybutyrate) (PHB), a raw material of biodegradable plastics and for production of hydrogen gas, and a renewable energy carrier by photosynthetic microorganisms (tentatively defined as cyanobacteria plus photosynthetic bateria, in this report). A thermophilic cyanobacterium, Synechococcus sp. MA19 that accumulates PHB at more than 20% of cell dry wt under nitrogen-starved conditions was isolated and microbiologically identified. The mechanism of PHB accumulation was studied. A mesophilic Synechococcus PCC7942 was transformed with the genes encoding PHB-synthesizing enzymes from Alcaligenes eutrophus. The transformant accumulated PHB under nitrogen-starved conditions. The optimal conditions for PHB accumulation by a photosynthetic bacterium grown on acetate were studied. Hydrogen production by photosynthetic microorganisms was studied. Cyanobacteria can produce hydrogen gas by nitrogenase or hydrogenase. Hydrogen production mediated by native hydrogenase in cyanobacteria was revealed to be in the dark anaerobic degradation of intracellular glycogen. A new system for light-dependent hydrogen production was targeted. In vitro and in vivo coupling of cyanobacterial ferredoxin with a heterologous hydrogenase was shown to produce hydrogen under light conditions. A trial for genetic trasformation of Synechococcus PCC7942 with the hydrogenase gene from Clostridium pasteurianum is going on. The strong hydrogen producers among photosynthetic bacteria were isolated and characterized. Co-culture of Rhodobacter and Clostriumdium was applied to produce hydrogen from glucose. Conversely in the case of cyanobacteria, genetic regulation of photosynthetic proteins was intended to improve conversion efficiency in hydrogen production by the photosynthetic bacterium, Rhodobacter sphaeroides RV. A mutant acquired by UV irradiation will be characterized for the mutation and for hydrogen productivity in comparison with the wild type strain. Some basic studies to develop photobioreactors are also introduced.

  5. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... production, production prices and production costs. 229.1204 Section 229.1204 Commodity and Securities... 1933, SECURITIES EXCHANGE ACT OF 1934 AND ENERGY POLICY AND CONSERVATION ACT OF 1975-REGULATION S-K... production, production prices and production costs. (a) For each of the last three fiscal years...

  6. 17 CFR 229.1204 - (Item 1204) Oil and gas production, production prices and production costs.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... production, production prices and production costs. 229.1204 Section 229.1204 Commodity and Securities... 1933, SECURITIES EXCHANGE ACT OF 1934 AND ENERGY POLICY AND CONSERVATION ACT OF 1975-REGULATION S-K... production, production prices and production costs. (a) For each of the last three fiscal years...

  7. Process for the production of fuel gas from coal

    DOEpatents

    Patel, Jitendra G. (Bolingbrook, IL); Sandstrom, William A. (Chicago, IL); Tarman, Paul B. (Elmhurst, IL)

    1982-01-01

    An improved apparatus and process for the conversion of hydrocarbonaceous materials, such as coal, to more valuable gaseous products in a fluidized bed gasification reaction and efficient withdrawal of agglomerated ash from the fluidized bed is disclosed. The improvements are obtained by introducing an oxygen containing gas into the bottom of the fluidized bed through a separate conduit positioned within the center of a nozzle adapted to agglomerate and withdraw the ash from the bottom of the fluidized bed. The conduit extends above the constricted center portion of the nozzle and preferably terminates within and does not extend from the nozzle. In addition to improving ash agglomeration and withdrawal, the present invention prevents sintering and clinkering of the ash in the fluidized bed and permits the efficient recycle of fine material recovered from the product gases by contacting the fines in the fluidized bed with the oxygen as it emanates from the conduit positioned within the withdrawal nozzle. Finally, the present method of oxygen introduction permits the efficient recycle of a portion of the product gases to the reaction zone to increase the reducing properties of the hot product gas.

  8. Fully automated radiochemical preparation system for gamma-spectroscopy on fission products and the study of the intruder and vibrational levels in /sup 83/Se

    SciTech Connect

    Lien, O.G. III

    1983-10-01

    AUTOBATCH was developed to provide a usable source of short-lived neutron-rich nuclides through chemical preparation of the sample from fission products for detailed gamma-ray spectroscopy, which would complement the output of on-line isotope separators. With AUTOBATCH the gamma rays following the ..beta../sup -/ decay of /sup 83/As were studied to determine the ground state spin and parity of /sup 83/As to be 5/2/sup -/; the absolute intensity of the ..beta../sup -/ branch from /sup 83/As to /sup 83/Se/sup m/ to be 0.3%; the absolute intensity of the ground state ..beta../sup -/ branch from /sup 83/Se/sup m/ to /sup 83/Br to be 39%; the halflife of the 5/2/sub 1//sup +/ level to be 3.2 ns; and the structure of /sup 83/Se/sub 49/. Results are used to show that the intruder structure which had been previously observed in the odd mass /sub 49/In isotopes could be observed in the N = 49 isotones. The observed structure is discussed in terms of the unified model calculations of Heyde which has been used to describe the intruder structure in the indium nuclei. The intruder structure is most strongly developed, not at core mid-shell, /sup 89/Zr/sub 49/, but rather at core mid-sub-shell /sup 83/Se. This difference is qualitatively understood to be due to the blocking of collectivity by the Z = 40 subshell closure which prevents the intruder structure from occurring in /sup 87/Sr/sub 49/ and /sup 89/Zr/sub 49/.

  9. Multi-phase glass-ceramics as a waste form for combined fission products: alkalis, alkaline earths, lanthanides, and transition metals

    SciTech Connect

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

    2012-04-01

    In this study, multi-phase silicate-based glass-ceramics were investigated as an alternate waste form for immobilizing non-fissionable products from used nuclear fuel. Currently, borosilicate glass is the waste form selected for immobilization of this waste stream, however, the low thermal stability and solubility of MoO{sub 3} in borosilicate glass translates into a maximum waste loading in the range of 15-20 mass%. Glass-ceramics provide the opportunity to target durable crystalline phases, e.g., powellite, oxyapatite, celsian, and pollucite, that will incorporate MoO{sub 3} as well as other waste components such as lanthanides, alkalis, and alkaline earths at levels 2X the solubility limits of a single-phase glass. In addition a glass-ceramic could provide higher thermal stability, depending upon the properties of the crystalline and amorphous phases. Glass-ceramics were successfully synthesized at waste loadings of 42, 45, and 50 mass% with the following glass additives: B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, CaO and SiO{sub 2} by slow cooling form from a glass melt. Glass-ceramics were characterized in terms of phase assemblage, morphology, and thermal stability. The targeted phases: powellite and oxyapatite were observed in all of the compositions along with a lanthanide borosilicate, and cerianite. Results of this initial investigation of glass-ceramics show promise as a potential waste form to replace single-phase borosilicate glass.

  10. Frack Attack: Weighing the Debate over the Hazards of Shale Gas Production

    E-print Network

    Frack Attack: Weighing the Debate over the Hazards of Shale Gas Production to be a "cleaner" fossil fuel with fewer, less potent emissions. However, shale gas horizons, it is possible that extensive methane emissions from shale gas operations

  11. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  12. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  13. 30 CFR 1202.550 - How do I determine the royalty due on gas production?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...do I determine the royalty due on gas production? 1202.550 Section... Mineral Resources OFFICE OF NATURAL RESOURCES REVENUE, DEPARTMENT OF THE INTERIOR NATURAL RESOURCES REVENUE ROYALTIES Gas Production From Indian Leases...

  14. Improved analytical capacity for determination of forage quality, utilising the gas production technique

    E-print Network

    Improved analytical capacity for determination of forage quality, utilising the gas production of Agricultural Research for Northern Sweden Rapport 2:2008 #12;#12;Contents Page ABSTRACT method.............................................................. 12 THE GAS PRODUCTION TECHNIQUE

  15. Improved Calculation of Thermal Fission Energy

    E-print Network

    X. B. Ma; W. L. Zhong; L. Z. Wang; Y. X. Chen; J. Cao

    2013-06-30

    Thermal fission energy is one of the basic parameters needed in the calculation of antineutrino flux for reactor neutrino experiments. It is useful to improve the precision of the thermal fission energy calculation for current and future reactor neutrino experiments, which are aimed at more precise determination of neutrino oscillation parameters. In this article, we give new values for thermal fission energies of some common thermal reactor fuel isotopes, with improvements on three aspects. One is more recent input data acquired from updated nuclear databases. the second one is a consideration of the production yields of fission fragments from both thermal and fast incident neutrons for each of the four main fuel isotopes. The last one is more carefully calculation of the average energy taken away by antineutrinos in thermal fission with the comparison of antineutrino spectrum from different models. The change in calculated antineutrino flux due to the new values of thermal fission energy is about 0.32%, and the uncertainties of the new values are about 50% smaller.

  16. Critical Temperature from the Fission Data

    SciTech Connect

    Cherepanov, E. A.; Karnaukhov, V. A.

    2007-05-22

    Experimental and calculated data on the fission probability are compared for highly excited 188Os. The calculations have been made within the statistical model using the more reliable parameterizations for the temperature dependence or surface tension. It is concluded that the critical temperature for the nuclear liquid-gas phase transition is higher than 16 MeV.

  17. Production Characteristics of Oceanic Natural Gas Hydrate Reservoirs

    NASA Astrophysics Data System (ADS)

    Max, M. D.; Johnson, A. H.

    2014-12-01

    Oceanic natural gas hydrate (NGH) accumulations form when natural gas is trapped thermodynamically within the gas hydrate stability zone (GHSZ), which extends downward from the seafloor in open ocean depths greater than about 500 metres. As water depths increase, the thickness of the GHSZ thickens, but economic NGH deposits probably occur no deeper than 1 km below the seafloor. Natural gas (mostly methane) appears to emanate mostly from deeper sources and migrates into the GHSZ. The natural gas crystallizes as NGH when the pressure - temperature conditions within the GHSZ are reached and when the chemical condition of dissolved gas concentration in pore water is high enough to favor crystallization. Although NGH can form in both primary and secondary porosity, the principal economic target appears to be turbidite sands on deep continental margins. Because these are very similar to the hosts of more deeply buried conventional gas and oil deposits, industry knows how to explore for them. Recent improvements in a seismic geotechnical approach to NGH identification and valuation have been confirmed by drilling in the northern Gulf of Mexico and allow for widespread exploration for NGH deposits to begin. NGH concentrations occur in the same semi-consolidated sediments in GHSZs worldwide. This provides for a narrow exploration window with low acoustic attenuation. These sediments present the same range of relatively easy drilling conditions and formation pressures that are only slightly greater than at the seafloor and are essentially equalized by water in wellbores. Expensive conventional drilling equipment is not required. NGH is the only hydrocarbon that is stable at its formation pressures and incapable of converting to gas without artificial stimulation. We suggest that specialized, NGH-specific drilling capability will offer opportunities for much less expensive drilling, more complex wellbore layouts that improve reservoir connectivity and in which gas-water separation can begin within the seafloor, and specialized production techniques. NGH is the only oceanic hydrocarbon deposit in which pressure can be controlled within the reservoir by balancing conversion and extraction. Oceanic NGH has a very low environmental risk, which also serves to distinguish it from other deepwater hydrocarbon deposits.

  18. Nanopowder production by gas-embedded electrical explosion of wire

    NASA Astrophysics Data System (ADS)

    Zou, Xiao-Bing; Mao, Zhi-Guo; Wang, Xin-Xin; Jiang, Wei-Hua

    2013-04-01

    A small electrical explosion of wire (EEW) setup for nanopowder production is constructed. It consists of a low inductance capacitor bank of 2 ?F-4 ?F typically charged to 8 kV-30 kV, a triggered gas switch, and a production chamber housing the exploding wire load and ambient gas. With the EEW device, nanosize powders of titanium oxides, titanium nitrides, copper oxides, and zinc oxides are successfully synthesized. The average particle size of synthesized powders under different experimental conditions is in a range of 20 nm-80 nm. The pressure of ambient gas or wire vapor can strongly affect the average particle size. The lower the pressure, the smaller the particle size is. For wire material with relatively high resistivity, such as titanium, whose deposited energy Wd is often less than sublimation energy Ws due to the flashover breakdown along the wire prematurely ending the Joule heating process, the synthesized particle size of titanium oxides or titanium nitrides increases with overheat coefficient k (k = Wd/Ws) increasing.

  19. Electron-capture delayed fission properties of 244Es

    SciTech Connect

    Shaughnessy, Dawn A.; Gregorich, Kenneth E.; Adams, Jeb L.; Lane, Michael R.; Laue, Carola A.; Lee, Diana M.; McGrath, Christopher A.; Ninov, Victor; Patin, Joshua B.; Strellis, Dan A.; Sylwester, Eric R.; Wilk, Philip A.; Hoffman, Darleane C.

    2001-03-16

    Electron-capture delayed fission was observed in {sup 244}Es produced via the {sup 237}Np({sup 12}C,5n){sup 244}Es reaction at 81 MeV (on target) with a production cross section of 0.31{+-}0.12 {micro}b. The mass-yield distribution of the fission fragments is highly asymmetric. The average preneutron-emission total kinetic energy of the fragments was measured to be 186{+-}19 MeV. Based on the ratio of the number of fission events to the measured number of {alpha} decays from the electron-capture daughter {sup 244}Cf (100% {alpha} branch), the probability of delayed fission was determined to be (1.2{+-}0.4) x 10{sup -4}. This value for the delayed fission probability fits the experimentally observed trend of increasing delayed fission probability with increasing Q value for electron-capture.

  20. The VERDI fission fragment spectrometer

    NASA Astrophysics Data System (ADS)

    Frégeau, M. O.; Bry?, T.; Gamboni, Th.; Geerts, W.; Oberstedt, S.; Oberstedt, A.; Borcea, R.

    2013-12-01

    The VERDI time-of-flight spectrometer is dedicated to measurements of fission product yields and of prompt neutron emission data. Pre-neutron fission-fragment masses will be determined by the double time-of-flight (TOF) technique. For this purpose an excellent time resolution is required. The time of flight of the fragments will be measured by electrostatic mirrors located near the target and the time signal coming from silicon detectors located at 50 cm on both sides of the target. This configuration, where the stop detector will provide us simultaneously with the kinetic energy of the fragment and timing information, significantly limits energy straggling in comparison to legacy experimental setup where a thin foil was usually used as a stop detector. In order to improve timing resolution, neutron transmutation doped silicon will be used. The high resistivity homogeneity of this material should significantly improve resolution in comparison to standard silicon detectors. Post-neutron fission fragment masses are obtained form the time-of-flight and the energy signal in the silicon detector. As an intermediary step a diamond detector will also be used as start detector located very close to the target. Previous tests have shown that poly-crystalline chemical vapour deposition (pCVD) diamonds provides a coincidence time resolution of 150 ps not allowing complete separation between very low-energy fission fragments, alpha particles and noise. New results from using artificial single-crystal diamonds (sCVD) show similar time resolution as from pCVD diamonds but also sufficiently good energy resolution.

  1. Production of biofuels from synthesis gas using microbial catalysts.

    PubMed

    Tirado-Acevedo, Oscar; Chinn, Mari S; Grunden, Amy M

    2010-01-01

    World energy consumption is expected to increase 44% in the next 20 years. Today, the main sources of energy are oil, coal, and natural gas, all fossil fuels. These fuels are unsustainable and contribute to environmental pollution. Biofuels are a promising source of sustainable energy. Feedstocks for biofuels used today such as grain starch are expensive and compete with food markets. Lignocellulosic biomass is abundant and readily available from a variety of sources, for example, energy crops and agricultural/industrial waste. Conversion of these materials to biofuels by microorganisms through direct hydrolysis and fermentation can be challenging. Alternatively, biomass can be converted to synthesis gas through gasification and transformed to fuels using chemical catalysts. Chemical conversion of synthesis gas components can be expensive and highly susceptible to catalyst poisoning, limiting biofuel yields. However, there are microorganisms that can convert the CO, H(2), and CO(2) in synthesis gas to fuels such as ethanol, butanol, and hydrogen. Biomass gasification-biosynthesis processing systems have shown promise as some companies have already been exploiting capable organisms for commercial purposes. The discovery of novel organisms capable of higher product yield, as well as metabolic engineering of existing microbial catalysts, makes this technology a viable option for reducing our dependency on fossil fuels. PMID:20359454

  2. 30 CFR 260.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 2 2011-07-01 2011-07-01 false How do I measure natural gas production on my... Bidding Systems Eligible Leases § 260.116 How do I measure natural gas production on my eligible lease? You must measure natural gas production on your eligible lease subject to the royalty...

  3. An automated system for measuring gas production from forages inoculated with rumen uid and its use

    E-print Network

    Griffith, Gareth

    inoculated with rumen ¯uid. The design of the apparatus and its method of use enables gas production of different systems have been used to measure gas production. Menke et al. (1979) described a method in whichAn automated system for measuring gas production from forages inoculated with rumen ¯uid and its

  4. First fission mass yield measurements using SPIDER at LANSCE

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

    Robust measurements of fission product properties, including mass yields, are important for advancing our understanding of the complex fission process and as improved inputs to calculation and simulation efforts in nuclear applications. The SPIDER detector, located at the Los Alamos Neutron Science Center (LANSCE), is a recently developed mass spectrometer aimed at measuring fission product mass yields with high resolution as a function of incident neutron energy and product mass, charge, and kinetic energy. The prototype SPIDER detector has been assembled, tested, installed at the Lujan Center at LANSCE, and taken initial thermal neutron induced measurements. The first results of mass yields for spontaneous fission of 252Cf and thermal neutron-induced fission of 235U measured with SPIDER will be presented. Ongoing upgrades and future plans for SPIDER will also be discussed. This work is in part supported by LANL Laboratory Directed Research and Development Projects 20110037DR and 20120077DR. LA-UR-14-24830.

  5. Development of temporary subtropical wetlands induces higher gas production

    PubMed Central

    Canterle, Eliete B.; da Motta Marques, David; Rodrigues, Lúcia R.

    2013-01-01

    Temporary wetlands are short-term alternative ecosystems formed by flooding for irrigation of areas used for rice farming. The goal of this study is to describe the development cycle of rice fields as temporary wetlands in southern Brazil, evaluating how this process affect the gas production (CH4 and CO2) in soil with difference % carbon and organic matter content. Two areas adjacent to Lake Mangueira in southern Brazil were used during a rice-farming cycle. One area had soil containing 1.1% carbon and 2.4% organic matter, and the second area had soil with 2.4% carbon and 4.4% organic matter. The mean rates of gas production were 0.04 ± 0.02 mg CH4 m?2 d?1 and 1.18 ± 0.30 mg CO2 m?2 d?1 in the soil area with the lower carbon content, and 0.02 ± 0.03 mg CH4 m?2 d?1 and 1.38 ± 0.41 mg CO2 m?2 d?1 in the soil area with higher carbon content. Our results showed that mean rates of CO2 production were higher than those of CH4 in both areas. No statistically significant difference was observed for production of CH4 considering different periods and sites. For carbon dioxide (CO2), however, a Two-Way ANOVA showed statistically significant difference (p = 0.05) considering sampling time, but no difference between areas. The results obtained suggest that the carbon and organic matter contents in the soil of irrigated rice cultivation areas may have been used in different ways by soil microorganisms, leading to variations in CH4 and CO2 production. PMID:23508352

  6. Fission Systems for Mars Exploration

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Kim, T.; Dorney, D. J.; Swint, Marion Shayne

    2012-01-01

    Fission systems are used extensively on earth, and 34 such systems have flown in space. The energy density of fission is over 10 million times that of chemical reactions, giving fission the potential to eliminate energy density constraints for many space missions. Potential safety and operational concerns with fission systems are well understood, and strategies exist for affordably developing such systems. By enabling a power-rich environment and highly efficient propulsion, fission systems could enable affordable, sustainable exploration of Mars.

  7. Singlet fission photovoltaics

    E-print Network

    Lee, Jiye

    2013-01-01

    The efficiency of a solar cell is restricted by the "single junction limit," whereby photons with energy higher than the bandgap lose energy by thermalization. Singlet exciton fission splits a high-energy molecular excitation ...

  8. Trash to Gas: Converting Space Waste into Useful Supply Products

    NASA Technical Reports Server (NTRS)

    Tsoras, Alexandra

    2013-01-01

    The cost of sending mass into space with current propulsion technology is very expensive, making every item a crucial element of the space mission. It is essential that all materials be used to their fullest potential. Items like food, packaging, clothing, paper towels, gloves, etc., normally become trash and take up space after use. These waste materials are currently either burned up upon reentry in earth's atmosphere or sent on cargo return vehicles back to earth: a very wasteful method. The purpose of this project was to utilize these materials and create useful products like water and methane gas, which is used for rocket fuel, to further supply a deep space mission. The system used was a thermal degradation reactor with the configuration of a down-draft gasifier. The reactor was loaded with approximately 100g of trash simulant and heated with two external ceramic heaters with separate temperature control in order to create pyrolysis and gasification in one zone and incineration iri a second zone simultaneously. Trash was loaded into the top half of the reactor to undergo pyrolysis while the downdraft gas experienced gasification or incineration to treat tars and maximize the production of carbon dioxide. Minor products included carbon monoxide, methane, and other hydrocarbons. The carbon dioxide produced can be sent to a Sabatier reactor to convert the gas into methane, which can be used as rocket propellant. In order to maximize the carbon dioxide and useful gases produced, and minimize the unwanted tars and leftover ashen material, multiple experiments were performed with altered parameters such as differing temperatures, flow rates, and location of inlet air flow. According to the data received from these experiments, the process will be further scaled up and optimized to ultimately create a system that reduces trash buildup while at the same time providing enough useful gases to potentially fill a methane tank that could fuel a lunar ascent vehicle or other deep space mission.

  9. 78 FR 52239 - Oil and Gas and Sulphur Operations on the Outer Continental Shelf-Oil and Gas Production Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-22

    ... current regulations were originally published on April 1, 1988 (53 FR 10690). Since that time, various... Part 250 Oil and Gas and Sulphur Operations on the Outer Continental Shelf--Oil and Gas Production... Part 250 RIN 1014-AA10 Oil and Gas and Sulphur Operations on the Outer Continental Shelf--Oil and...

  10. Singlet exciton fission photovoltaics.

    PubMed

    Lee, Jiye; Jadhav, Priya; Reusswig, Philip D; Yost, Shane R; Thompson, Nicholas J; Congreve, Daniel N; Hontz, Eric; Van Voorhis, Troy; Baldo, Marc A

    2013-06-18

    Singlet exciton fission, a process that generates two excitons from a single photon, is perhaps the most efficient of the various multiexciton-generation processes studied to date, offering the potential to increase the efficiency of solar devices. But its unique characteristic, splitting a photogenerated singlet exciton into two dark triplet states, means that the empty absorption region between the singlet and triplet excitons must be filled by adding another material that captures low-energy photons. This has required the development of specialized device architectures. In this Account, we review work to develop devices that harness the theoretical benefits of singlet exciton fission. First, we discuss singlet fission in the archetypal material, pentacene. Pentacene-based photovoltaic devices typically show high external and internal quantum efficiencies. They have enabled researchers to characterize fission, including yield and the impact of competing loss processes, within functional devices. We review in situ probes of singlet fission that modulate the photocurrent using a magnetic field. We also summarize studies of the dissociation of triplet excitons into charge at the pentacene-buckyball (C60) donor-acceptor interface. Multiple independent measurements confirm that pentacene triplet excitons can dissociate at the C60 interface despite their relatively low energy. Because triplet excitons produced by singlet fission each have no more than half the energy of the original photoexcitation, they limit the potential open circuit voltage within a solar cell. Thus, if singlet fission is to increase the overall efficiency of a solar cell and not just double the photocurrent at the cost of halving the voltage, it is necessary to also harvest photons in the absorption gap between the singlet and triplet energies of the singlet fission material. We review two device architectures that attempt this using long-wavelength materials: a three-layer structure that uses long- and short-wavelength donors and an acceptor and a simpler, two-layer combination of a singlet-fission donor and a long-wavelength acceptor. An example of the trilayer structure is singlet fission in tetracene with copper phthalocyanine inserted at the C60 interface. The bilayer approach includes pentacene photovoltaic cells with an acceptor of infrared-absorbing lead sulfide or lead selenide nanocrystals. Lead selenide nanocrystals appear to be the most promising acceptors, exhibiting efficient triplet exciton dissociation and high power conversion efficiency. Finally, we review architectures that use singlet fission materials to sensitize other absorbers, thereby effectively converting conventional donor materials to singlet fission dyes. In these devices, photoexcitation occurs in a particular molecule and then energy is transferred to a singlet fission dye where the fission occurs. For example, rubrene inserted between a donor and an acceptor decouples the ability to perform singlet fission from other major photovoltaic properties such as light absorption. PMID:23611026

  11. Agricultural use of a flue gas desulfurization by-product

    SciTech Connect

    Nelson, S. Jr.; Dick, W.; Chen, L.

    1998-04-01

    Few, if any, economical alternatives exist for operators of small coal-fired boilers that require a flue-gas desulfurization system which does not generate wastes. A new duct-injection technology called {open_quotes}Fluesorbent{close_quotes} has been developed to help fill this gap. Fluesorbent FGD was intentionally designed so that the saturated SO{sub 2}-sorbent materials would be valuable soil amendments for agricultural or turf-grass land. Agricultural and turf grass studies recently commenced using spent Fluesorbent materials from an FGD pilot program at an Ohio power plant. In the first year of testing, alfalfa yields on field plots with the FGD by-products were approximately 250% greater than on plots with no treatment, and about 40% greater than on plots treated with an equivalent amount of agricultural lime. Because the FGD by-products contained trace elements from included fly ash, the chemical composition of the alfalfa was significantly improved.

  12. Agricultural use of a flue gas desulfurization by-product

    SciTech Connect

    Nelson, S. Jr.; Dick, W.; Chen, L.

    1998-07-01

    Few, if any, economical alternatives exist for operators of small coal-fired boilers that require a flue-gas desulfurization system which does not generate wastes. A new duct-injection technology called Fluesorbent has been developed to help fill this gap. Fluesorbent FGD was intentionally designed so that the saturated SO{sub 2}-sorbent materials would be valuable solid amendments for agricultural or turf-grass land. Agricultural and turf grass studies recently commenced using spent Fluesorbent materials from an FGD pilot program at an Ohio power plant. In the first year of testing, alfalfa yields on field plots with the FGS by-products were approximately 250% greater than on plots with no treatment, and about 40% greater than on plots treated with an equivalent amount of agricultural lime. Because the FGD by-products contained trace elements from included fly ash, the chemical composition of the alfalfa was significantly improved.

  13. Agricultural use of a flue gas desulfurization by-product

    SciTech Connect

    Dick, W.; Chen, L.; Nelson, S. Jr.

    1998-12-31

    Few, if any, economical alternatives exist for operators of small coal-fired boilers that require a flue-gas desulfurization system which does not generate wastes. A new duct-injection technology called Fluesorbent has been developed to help fill this gap. Fluesorbent FGD was intentionally designed so that the saturated SO{sub 2}-sorbent materials would be valuable soil amendments for agricultural or turf-grass land. Agricultural and turf grass studies recently commenced using spent Fluesorbent materials from an FGD pilot program at an Ohio power plant. In the first year of testing, alfalfa yields on field plots with the FGD by-products were approximately 250% greater than on plots with no treatment, and about 40% greater than on plots treated with an equivalent amount of agricultural lime. Because the FGD by-products contained trace elements from included fly ash, the chemical composition of the alfalfa was significantly improved. Detailed yield and chemical data are presented.

  14. Application of Fission Chamber to Uranium Microanalysis

    NASA Astrophysics Data System (ADS)

    Yamada, Kimio; Izumi, Shigeru; Otsuka, Hisao; Matsumoto, Tetsuo

    1983-02-01

    The possibility of using a fission chamber for quantitative analysis of uranium impurity in dynamic memory materials was studied. The fission chamber had two pairs of parallel disk electrodes. One electrode of each pair was used as a collector and was made of Teflon with a pure aluminum coating, while the other electrode was the material to be measured. Carbon dioxide was used as the ionization gas. Uranium in the materials was irradiated with neutrons and the number of fissions was counted to give the impurity content. Uranium contents in aluminum (99.8%) and Teflon were calculated, and measured values showed a fairly good reproducibility. The detection limit, determined by background fluctuations, for uranium impurity contained in the aluminum coated Teflon electrode was 4.0 ppb.

  15. Air quality concerns of unconventional oil and natural gas production.

    PubMed

    Field, R A; Soltis, J; Murphy, S

    2014-05-01

    Increased use of hydraulic fracturing ("fracking") in unconventional oil and natural gas (O & NG) development from coal, sandstone, and shale deposits in the United States (US) has created environmental concerns over water and air quality impacts. In this perspective we focus on how the production of unconventional O & NG affects air quality. We pay particular attention to shale gas as this type of development has transformed natural gas production in the US and is set to become important in the rest of the world. A variety of potential emission sources can be spread over tens of thousands of acres of a production area and this complicates assessment of local and regional air quality impacts. We outline upstream activities including drilling, completion and production. After contrasting the context for development activities in the US and Europe we explore the use of inventories for determining air emissions. Location and scale of analysis is important, as O & NG production emissions in some US basins account for nearly 100% of the pollution burden, whereas in other basins these activities make up less than 10% of total air emissions. While emission inventories are beneficial to quantifying air emissions from a particular source category, they do have limitations when determining air quality impacts from a large area. Air monitoring is essential, not only to validate inventories, but also to measure impacts. We describe the use of measurements, including ground-based mobile monitoring, network stations, airborne, and satellite platforms for measuring air quality impacts. We identify nitrogen oxides, volatile organic compounds (VOC), ozone, hazardous air pollutants (HAP), and methane as pollutants of concern related to O & NG activities. These pollutants can contribute to air quality concerns and they may be regulated in ambient air, due to human health or climate forcing concerns. Close to well pads, emissions are concentrated and exposure to a wide range of pollutants is possible. Public health protection is improved when emissions are controlled and facilities are located away from where people live. Based on lessons learned in the US we outline an approach for future unconventional O & NG development that includes regulation, assessment and monitoring. PMID:24699994

  16. Characterizing tight-gas systems with production data: Wyoming, Utah, and Colorado

    USGS Publications Warehouse

    Nelson, Philip H.; Santus, Stephen L.

    2013-01-01

    The study of produced fluids allows comparisons among tight-gas systems. This paper examines gas, oil, and water production data from vertical wells in 23 fields in five Rocky Mountain basins of the United States, mostly from wells completed before the year 2000. Average daily rates of gas, oil, and water production are determined two years and seven years after production begins in order to represent the interval in which gas production declines exponentially. In addition to the daily rates, results are also presented in terms of oil-to-gas and water-to-gas ratios, and in terms of the five-year decline in gas production rates and water-to-gas ratios. No attempt has been made to estimate the ultimate productivity of wells or fields. The ratio of gas production rates after seven years to gas production rates at two years is about one-half, with median ratios falling within a range of 0.4 to 0.6 in 16 fields. Oil-gas ratios show substantial variation among fields, ranging from dry gas (no oil) to wet gas to retrograde conditions. Among wells within fields, the oil-gas ratios vary by a factor of three to thirty, with the exception of the Lance Formation in Jonah and Pinedale fields, where the oil-gas ratios vary by less than a factor of two. One field produces water-free gas and a large fraction of wells in two other fields produce water-free gas, but most fields have water-gas ratios greater than 1 bbl/mmcf—greater than can be attributed to water dissolved in gas in the reservoir— and as high as 100 bbl/mmcf. The median water-gas ratio for fields increases moderately with time, but in individual wells water influx relative to gas is erratic, increasing greatly with time in many wells while remaining constant or decreasing in others.

  17. CCl bond fission dynamics and angular momentum recoupling in the 235 nm photodissociation of allyl chloride

    E-print Network

    Butler, Laurie J.

    C­Cl bond fission dynamics and angular momentum recoupling in the 235 nm photodissociation of allyl/2) products by 2 1 resonance enhanced multiphoton ionization shows that primary C­Cl bond fission of allyl(2 P3/2) of 0.097/0.903. The minor dissociation channel for C­Cl bond fission, producing low kinetic

  18. Microbiology of synthesis gas fermentation for biofuel production.

    PubMed

    Henstra, Anne M; Sipma, Jan; Rinzema, Arjen; Stams, Alfons J M

    2007-06-01

    A significant portion of biomass sources like straw and wood is poorly degradable and cannot be converted to biofuels by microorganisms. The gasification of this waste material to produce synthesis gas (or syngas) could offer a solution to this problem, as microorganisms that convert CO and H2) (the essential components of syngas) to multicarbon compounds are available. These are predominantly mesophilic microorganisms that produce short-chain fatty acids and alcohols from CO and H2. Additionally, hydrogen can be produced by carboxydotrophic hydrogenogenic bacteria that convert CO and H2O to H2 and CO2. The production of ethanol through syngas fermentation is already available as a commercial process. The use of thermophilic microorganisms for these processes could offer some advantages; however, to date, few thermophiles are known that grow well on syngas and produce organic compounds. The identification of new isolates that would broaden the product range of syngas fermentations is desirable. Metabolic engineering could be employed to broaden the variety of available products, although genetic tools for such engineering are currently unavailable. Nevertheless, syngas fermenting microorganisms possess advantageous characteristics for biofuel production and hold potential for future engineering efforts. PMID:17399976

  19. Greenhouse gas emission associated with sugar production in southern Brazil

    PubMed Central

    2010-01-01

    Background Since sugarcane areas have increased rapidly in Brazil, the contribution of the sugarcane production, and, especially, of the sugarcane harvest system to the greenhouse gas emissions of the country is an issue of national concern. Here we analyze some data characterizing various activities of two sugarcane mills during the harvest period of 2006-2007 and quantify the carbon footprint of sugar production. Results According to our calculations, 241 kg of carbon dioxide equivalent were released to the atmosphere per a ton of sugar produced (2406 kg of carbon dioxide equivalent per a hectare of the cropped area, and 26.5 kg of carbon dioxide equivalent per a ton of sugarcane processed). The major part of the total emission (44%) resulted from residues burning; about 20% resulted from the use of synthetic fertilizers, and about 18% from fossil fuel combustion. Conclusions The results of this study suggest that the most important reduction in greenhouse gas emissions from sugarcane areas could be achieved by switching to a green harvest system, that is, to harvesting without burning. PMID:20565736

  20. The Economic Impact of Shale Gas Production in the U.S

    NASA Astrophysics Data System (ADS)

    Yang, Yang

    Energy is important to our daily lives. A price change of one energy type may influence our consumption choices, commodities prices and industry production. For the United States, shale gas is becoming a promising source of natural gas because of the rapid increase in its reserve and production capacity. Shale gas production is projected to be a large proportion of U.S. gas production, as predicted by Energy Information Administration (EIA). However, besides knowing the big picture, more details are needed before characterizing shale gas as a "game changer." It is interesting to address questions like to what extent the production of shale gas could affect other industries' production, stabilize commodities' prices, and what are the impacts on factor payments, capital returns, labor payments and household consumption. In this study, I use a CGE model to measure the impact on industry and the change in social welfare associated with shale gas production.

  1. Storage sizing for embedding of local gas production in a micro gas grid

    NASA Astrophysics Data System (ADS)

    Alkano, D.; Nefkens, W. J.; Scherpen, J. M. A.; Volkerts, M.

    2014-12-01

    In this paper we study the optimal control of a micro grid of biogas producers. The paper considers the possibility to have a local storage device for each producer, who partly consumes his own production, i.e. prosumer. In addition, connected prosumers can sell stored gas to create revenue from it. An optimization model is employed to derive the size of storage device and to provide a pricing mechanism in an effort to value the stored gas. Taking into account physical grid constraints, the model is constructed in a centralized scheme of model predictive control. Case studies show that there is a relation between the demand and price profiles in terms of peaks and lows. The price profiles generally follow each other. The case studies are employed as well to to study the impacts of model parameters on deriving the storage size.

  2. Ground movements associated with gas hydrate production. Final report

    SciTech Connect

    Siriwardane, H.J.; Kutuk, B.

    1992-03-01

    This report deals with a study directed towards a modeling effort on production related ground movements and subsidence resulting from hydrate dissociation. The goal of this research study was to evaluate whether there could be subsidence related problems that could be an impediment to hydrate production. During the production of gas from a hydrate reservoir, it is expected that porous reservoir matrix becomes more compressible which may cause reservoir compression (compaction) under the influence of overburden weight. The overburden deformations can propagate its influence upwards causing subsidence near the surface where production equipment will be located. In the present study, the reservoir compaction is modeled by using the conventional ``stress equilibrium`` approach. In this approach, the overburden strata move under the influence of body force (i.e. self weight) in response to the ``cavity`` generated by reservoir depletion. The present study is expected to provide a ``lower bound`` solution to the subsidence caused by hydrate reservoir depletion. The reservoir compaction anticipated during hydrate production was modeled by using the finite element method, which is a powerful computer modeling technique. The ground movements at the reservoir roof (i.e. reservoir compression) cause additional stresses and disturbance in the overburden strata. In this study, the reservoir compaction was modeled by using the conventional ``stress equilibrium`` approach. In this approach, the overburden strata move under the influence of body force (i.e. self weight) in response to the ``cavity`` generated by reservoir depletion. The resulting stresses and ground movements were computed by using the finite element method. Based on the parameters used in this investigation, the maximum ground subsidence could vary anywhere from 0.50 to 6.50 inches depending on the overburden depth and the size of the depleted hydrate reservoir.

  3. Fission Surface Power Technology Development Update

    NASA Technical Reports Server (NTRS)

    Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

    2011-01-01

    Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

  4. Future Scenarios for Fission Based Reactors

    NASA Astrophysics Data System (ADS)

    David, S.

    2005-04-01

    The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile inventory, capacity to be deployed, induced radiotoxicities, and R&D efforts.

  5. Fission neutron spectra measurements at LANSCE - status and plans

    SciTech Connect

    Haight, Robert C; Noda, Shusaku; Nelson, Ronald O; O' Donnell, John M; Devlin, Matt; Chatillon, Audrey; Granier, Thierry; Taieb, Julien; Laurent, Benoit; Belier, Gilbert; Becker, John A; Wu, Ching - Yen

    2009-01-01

    A program to measure fission neutron spectra from neutron-induced fission of actinides is underway at the Los Alamos Neutron Science Center (LANSCE) in a collaboration among the CEA laboratory at Bruyeres-le-Chatel, Lawrence Livermore National Laboratory and Los Alamos National Laboratory. The spallation source of fast neutrons at LANSCE is used to provide incident neutron energies from less than 1 MeV to 100 MeV or higher. The fission events take place in a gas-ionization fission chamber, and the time of flight from the neutron source to that chamber gives the energy of the incident neutron. Outgoing neutrons are detected by an array of organic liquid scintillator neutron detectors, and their energies are deduced from the time of flight from the fission chamber to the neutron detector. Measurements have been made of the fission neutrons from fission of {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. The range of outgoing energies measured so far is from 1 MeV to approximately 8 MeV. These partial spectra and average fission neutron energies are compared with evaluated data and with models of fission neutron emission. Results to date will be presented and a discussion of uncertainties will be given in this presentation. Future plans are to make significant improvements in the fission chambers, neutron detectors, signal processing, data acquisition and the experimental environment to provide high fidelity data including mea urements of fission neutrons below 1 MeV and improvements in the data above 8 MeV.

  6. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... feet of natural gas, measured according to 30 CFR part 250, subpart L, equals one barrel of oil... 30 Mineral Resources 2 2013-07-01 2013-07-01 false How do I measure natural gas production on my... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural...

  7. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... feet of natural gas, measured according to 30 CFR part 250, subpart L, equals one barrel of oil... 30 Mineral Resources 2 2012-07-01 2012-07-01 false How do I measure natural gas production on my... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural...

  8. 30 CFR 560.116 - How do I measure natural gas production on my eligible lease?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... feet of natural gas, measured according to 30 CFR part 250, subpart L, equals one barrel of oil... 30 Mineral Resources 2 2014-07-01 2014-07-01 false How do I measure natural gas production on my... § 560.116 How do I measure natural gas production on my eligible lease? You must measure natural...

  9. Liquid uranium alloy-helium fission reactor

    DOEpatents

    Minkov, V.

    1984-06-13

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  10. Liquid uranium alloy-helium fission reactor

    DOEpatents

    Minkov, Vladimir (Skokie, IL)

    1986-01-01

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

  11. 78 FR 59632 - Oil and Gas and Sulphur Operations on the Outer Continental Shelf-Oil and Gas Production Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-27

    ... Operations on the Outer Continental Shelf--Oil and Gas Production Safety Systems AGENCY: Bureau of Safety and... proposed rulemaking on production safety systems on August 22, 2013 (78 FR 52240). The proposed rule would... operating dry tree and subsea tree production systems on the Outer Continental Shelf (OCS) and divide...

  12. BIOENGINEERING, FOOD, AND NATURAL PRODUCTS Gas Permeance Measurement of Hollow Fiber

    E-print Network

    Federspiel, William J.

    BIOENGINEERING, FOOD, AND NATURAL PRODUCTS Gas Permeance Measurement of Hollow Fiber Membranes in Gas-Liquid Environment Laura W. Lund McGowan Institute for Regenerative Medicine, Dept in a gas-liquid en®ironment. The gas-liquid permeance measurements were compared with measurements made

  13. An experimental investigation of 235 sub UF sub 6 fission produced plasmas. [gas handling system for use with nuclear pumped laser experiments

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1981-01-01

    A gas handling system capable of use with uranium fluoride was designed and constructed for use with nuclear pumped laser experiments using the TRIGA research reactor. By employing careful design and temperature controls, the UF6 can be first transported into the irradiation chamber, and then, at the conclusion of the experiment, returned to gas cylinders. The design of the system is described. Operating procedures for the UF6 and gas handling systems are included.

  14. Process for treating fission waste

    DOEpatents

    Rohrmann, Charles A. (Kennewick, WA); Wick, Oswald J. (Richland, WA)

    1983-01-01

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste.

  15. Study of extraterrestrial disposal of radioactive wastes. Part 3: Preliminary feasibility screening study of space disposal of the actinide radioactive wastes with 1 percent and 0.1 percent fission product contamination

    NASA Technical Reports Server (NTRS)

    Hyland, R. E.; Wohl, M. L.; Finnegan, P. M.

    1973-01-01

    A preliminary study was conducted of the feasibility of space disposal of the actinide class of radioactive waste material. This waste was assumed to contain 1 and 0.1 percent residual fission products, since it may not be feasible to completely separate the actinides. The actinides are a small fraction of the total waste but they remain radioactive much longer than the other wastes and must be isolated from human encounter for tens of thousands of years. Results indicate that space disposal is promising but more study is required, particularly in the area of safety. The minimum cost of space transportation would increase the consumer electric utility bill by the order of 1 percent for earth escape and 3 percent for solar escape. The waste package in this phase of the study was designed for normal operating conditions only; the design of next phase of the study will include provisions for accident safety. The number of shuttle launches per year required to dispose of all U.S. generated actinide waste with 0.1 percent residual fission products varies between 3 and 15 in 1985 and between 25 and 110 by 2000. The lower values assume earth escape (solar orbit) and the higher values are for escape from the solar system.

  16. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    SciTech Connect

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.

  17. Microscopic Description of Induced Fission

    E-print Network

    N. Schunck

    2013-02-22

    Selected aspects of the description of neutron-induced fission in 240Pu in the framework of the nuclear energy density functional theory at finite temperature are presented. In particular, we discuss aspects pertaining to the choice of thermodynamic state variables, the evolution of fission barriers as function of the incident neutron energy, and the temperatures of the fission fragments.

  18. Student Experiments in Spontaneous Fission.

    ERIC Educational Resources Information Center

    Becchetti, F. D.; Ying, J. S.

    1981-01-01

    Advanced undergraduate experiments utilizing a commercially available, thin spontaneous fission source are described, including studies of the energy and mass distribution of the fission fragments and their energy and angular correlation. The experiments provide a useful introduction to fission, nuclear mass equations, heavy-ion physics, and…

  19. Atmospheric emissions and air quality impacts from natural gas production and use.

    PubMed

    Allen, David T

    2014-01-01

    The US Energy Information Administration projects that hydraulic fracturing of shale formations will become a dominant source of domestic natural gas supply over the next several decades, transforming the energy landscape in the United States. However, the environmental impacts associated with fracking for shale gas have made it controversial. This review examines emissions and impacts of air pollutants associated with shale gas production and use. Emissions and impacts of greenhouse gases, photochemically active air pollutants, and toxic air pollutants are described. In addition to the direct atmospheric impacts of expanded natural gas production, indirect effects are also described. Widespread availability of shale gas can drive down natural gas prices, which, in turn, can impact the use patterns for natural gas. Natural gas production and use in electricity generation are used as a case study for examining these indirect consequences of expanded natural gas availability. PMID:24498952

  20. Modelling of fission chambers in current mode—Analytical approach

    NASA Astrophysics Data System (ADS)

    Chabod, Sébastien; Fioni, Gabriele; Letourneau, Alain; Marie, Frédéric

    2006-10-01

    A comprehensive theoretical model is proposed to explain the functioning of fission chambers operated in current mode, even in very high neutron fluxes. The calibration curves are calculated as a function of basic physical parameters as fission rate, gas pressure and geometry of the chambers. The output current at saturation is precisely calculated, as well as the maximum voltage to be applied in order to avoid avalanche phenomena. The electric field distortion due to the space charge phenomena is also estimated. Within this model, the characteristic responses of fission chambers are correctly reproduced, in agreement with the experience feedback obtained at the ILL/Grenoble High-Flux Reactor.