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1

A nuclide-separation wire precipitator for measurement of noble-gas fission products  

Microsoft Academic Search

A wire precipitator, which has a capability of determining separately the activity concentrations of three radioactive noble-gas fission products, 88Kr, 89Kr and 138Xe, was developed. The precipitator utilizes the characteristics that the respective daughter nuclides of the fission products emit beta particles of different energy spectra. The activity concentrations of the noble gas fission products were separately quantified even when

M. Katagiri; M. Kishimoto; H. Ito; H. Yoshida; M. Fukushima; H. Ohkawa; T. Saruta

1993-01-01

2

New Results on Helium and Tritium Gas Production From Ternary Fission  

SciTech Connect

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.

Serot, O. [CEA Cadarache, DEN/DER/SPRC/LEPh, F-13108 Saint Paul-Lez-Durance (France); Wagemans, C. [Dept. of Subatomic and Radiation Physics, University of Gent, B-9000 Gent (Belgium); Heyse, J. [EC-JRC-Institute for Reference Materials and Measurements, Retieseweg 111, B-2440 Geel (Belgium)

2005-05-24

3

Fission gas detection system  

DOEpatents

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.

Colburn, Richard P. (Pasco, WA)

1985-01-01

4

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

SciTech Connect

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.

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

2010-10-01

5

FIssion Product Prompt ?-ray spectrometer: Development of an instrumented gas-filled magnetic spectrometer at the ILL  

NASA Astrophysics Data System (ADS)

Accurate thermal neutron-induced fission data are important for applications in reactor physics as well as for fundamental nuclear physics. FIPPS is the new FIssion Product Prompt ?-ray Spectrometer being developed at the Institut Laue Langevin for neutron-induced fission studies. FIPPS is based on the combination of a large Germanium detector array surrounding a fission target, a Time-Of-Flight detector and a Gas-Filled Magnet (GFM) to identify mass, nuclear charge and kinetic energy of one of the fission fragments. The GFM will be instrumented with a Time-Projection Chamber (TPC) for individual 3D tracking of the fragments. A conceptual design study of the new spectrometer is presented.

Blanc, A.; Chebboubi, A.; Faust, H.; Jentschel, M.; Kessedjian, G.; Kster, U.; Materna, T.; Panebianco, S.; Sage, C.; Urban, W.

2013-12-01

6

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

SciTech Connect

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

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

2008-09-01

7

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

SciTech Connect

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

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

2010-10-01

8

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

NASA Technical Reports Server (NTRS)

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.

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

1974-01-01

9

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

NASA Astrophysics Data System (ADS)

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

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

2012-01-01

10

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

SciTech Connect

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

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

1987-01-01

11

Correlation of recent fission product release data  

SciTech Connect

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

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

1989-01-01

12

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

NASA Technical Reports Server (NTRS)

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

Capo, M. A.; Mickle, R.

1975-01-01

13

Antiproton Powered Gas Core Fission Rocket  

SciTech Connect

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.

Kammash, Terry [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States)

2005-02-06

14

Measurement of Fission Product Yields from Fast-Neutron Fission  

NASA Astrophysics Data System (ADS)

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.

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

15

Principles of a gas filled magnetic spectrometer for fission studies  

NASA Astrophysics Data System (ADS)

The spectroscopy of the prompt gamma decay from fission products gives information on the entry states, e.g. distribution functions for excitation energy and spin, and therefore a direct link to the fission process itself. This type of spectroscopy is, however, only possible when a filter can be constructed which allows setting a gate to the gamma-spectrum in a narrow region in mass and nuclear charge, as well as on the total excitation energy of the fragment split under investigation. A possible configuration of a prompt gamma-ray spectrometer consist of a gamma-ray array composed of high resolution germanium detectors, coupled to a gas filled magnet. We will outline the principles for a gas filled magnetic spectrometer for fission product spectroscopy. In particular the focusing characteristics of such a device, which are valid for particles in the velocity regime of E/A< 1MeV/amu, will be addressed. First experiments on the LOHENGRIN spectrometer in Grenoble investigating on the behavior of fission products in gas filled magnets have been performed, and have validated the experimental approach to the nuclear fission process with such a device.

Faust, H.; Chebboubi, A.; Kessedjian, G.; Sage, C.; Kster, U.; Blanc, A.

2013-12-01

16

Nondestructive fission gas release measurement and analysis  

Microsoft Academic Search

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

P. M. OLeary; D. R. Packard

1993-01-01

17

USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10  

SciTech Connect

This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.

NONE

1986-05-01

18

Reactor power history from fission product signatures  

E-print Network

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

Sweeney, David J.

2009-05-15

19

Modeling Fission Product Sorption in Graphite Structures  

SciTech Connect

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).

Szlufarska, Izabela [University of Wisconsin, Madison, WI (United States); Morgan, Dane [University of Wisconsin, Madison, WI (United States); Allen, Todd [University of Wisconsin, Madison, WI (United States)

2013-04-08

20

Recovery and use of fission product noble metals  

SciTech Connect

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)

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

1980-06-01

21

Systematics of Fission-Product Yields  

SciTech Connect

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.

A.C. Wahl

2002-05-01

22

Electron spectra from decay of fission products  

SciTech Connect

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.

Dickens, J K

1982-09-01

23

Fission product release from irradiated LWR fuel under accident conditions  

SciTech Connect

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

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

1984-01-01

24

Kinetics of fission product release prior to fuel slumping  

SciTech Connect

This paper describes the primary physical/chemical models recently incorporated into a mechanistic code (FASTGRASS) for the estimation of fission product release from fuel, and compares predicted results with test data. The theory of noble gas behavior is discussed in relation to its effect on the release behavior of I, Cs, Te, Ba, and Sr. The behavior of these fission products in the presence of fuel liquefaction/dissolution and oxidation grain-growth phenomena is presented, as is the chemistry of Sr, Ba, I, and Cs. Comparison of code predictions with data indicates the following trends. Fission product release behavior from solid strongly depends on fuel microstructure, irradiation history, time at temperature, and internal fuel rod chemistry. Fuel liquefaction/dissolution, fracturing, and oxidation also exert a pronounced effect on release during fuel rod degradation. For very low burnup fuel appreciable fission product retention in previously liquefied fuel can occur due to the low concentration of fission products, and the limited growth of bubbles in the liquefied material. 24 refs., 13 figs., 9 tabs.

Rest, J.

1987-10-01

25

THE INDUSTRIAL UTILISATION OF FISSION PRODUCTS  

Microsoft Academic Search

Industrial applications of fussion products are reviewed. In addition ; to the economic aspects of salvaging uranium fission products the increasing ; problem of storage will be partially solved. It is estimated that by 1975, 18.5 ; tons of U²³⁵ will be used in power plants of Great Britian alone each year. ; This would mean an almost equal mass

1955-01-01

26

Yields of fission products produced by thermal-neutron fission of 243Cm  

Microsoft Academic Search

On the basis of measured yields for 72 gamma rays and known nuclear data, cumulative fission-product yields were deduced for 69 fission products having half-lives between 36 seconds and 65 days representing 41 mass chains created during thermal-neutron fission of 243Cm.

J. K. Dickens; J. W. McConnell

1986-01-01

27

Yields of fission products produced by thermal-neutron fission of 243Cm  

NASA Astrophysics Data System (ADS)

On the basis of measured yields for 72 gamma rays and known nuclear data, cumulative fission-product yields were deduced for 69 fission products having half-lives between 36 seconds and 65 days representing 41 mass chains created during thermal-neutron fission of 243Cm.

Dickens, J. K.; McConnell, J. W.

1986-08-01

28

A fission gas release correlation for uranium nitride fuel pins  

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

29

Fission product yields for thermal-neutron fission of plutonium-239  

Microsoft Academic Search

Absolute cumulative yields have been determined for 49 fission products representing 36 mass chains created during thermal-neutron fission of ²³⁹Pu, including 3 mass chains for which no prior data exist. Using Ge(Li) spectroscopy, spectra were obtained of gamma rays from decay of fission products between 1550 s and 31 days after a 100-s irradiation. Data were obtained for all fission

J. K. Dickens; J. W. McConnell

1980-01-01

30

Fission gas release restrictor for breached fuel rod  

DOEpatents

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.

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

1986-01-01

31

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

32

Sensitivity analysis of the fission gas behavior model in BISON.  

SciTech Connect

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.

Swiler, Laura Painton; Pastore, Giovanni [Idaho National Laboratory, Idaho Fall, ID; Perez, Danielle [Idaho National Laboratory, Idaho Fall, ID; Williamson, Richard [Idaho National Laboratory, Idaho Fall, ID

2013-05-01

33

Fractal Model of Fission Product Release in Nuclear Fuel  

NASA Astrophysics Data System (ADS)

A model of fission gas migration in nuclear fuel pellet is proposed. Diffusion process of fission gas in granular structure of nuclear fuel with presence of inter-granular bubbles in the fuel matrix is simulated by fractional diffusion model. The Grunwald-Letnikov derivative parameter characterizes the influence of porous fuel matrix on the diffusion process of fission gas. A finite-difference method for solving fractional diffusion equations is considered. Numerical solution of diffusion equation shows correlation of fission gas release and Grunwald-Letnikov derivative parameter. Calculated profile of fission gas concentration distribution is similar to that obtained in the experimental studies. Diffusion of fission gas is modeled for real RBMK-1500 fuel operation conditions. A functional dependence of Grunwald-Letnikov derivative parameter with fuel burn-up is established.

Stankunas, Gediminas

2012-09-01

34

The behavior of fission products during nuclear rocket reactor tests  

SciTech Connect

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

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

1991-01-10

35

Fission and Nuclear Liquid-Gas Phase Transition  

E-print Network

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.

E. A. Cherepanov; V. A. Karnaukhov

2007-03-30

36

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

SciTech Connect

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

Rest, J.

1983-09-01

37

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

SciTech Connect

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

Rest, J.

1983-09-01

38

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

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

39

Modeling of Fission Gas Release in UO2  

SciTech Connect

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].

MH Krohn

2006-01-23

40

Energy Dependence of Plutonium Fission-Product Yields  

NASA Astrophysics Data System (ADS)

A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured 240Pu spontaneous fission and thermal-neutron-induced fission of 239Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + 239Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.

Lestone, J. P.

2011-12-01

41

Measurement of fission product gases in the atmosphere  

NASA Astrophysics Data System (ADS)

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

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

1997-01-01

42

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

SciTech Connect

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)

Fioni, G. [Commissariat Energie Atomique, Saclay (France); Faust, H.; Oberstedt, S. [Institut Laue-Langevin, Grenoble (France); Hambsch, F. [Institute for Reference Materials and Measurements, Geel (Belgium)

1998-10-01

43

Assessment of a mechanistic model in U-Pu-Zr metallic alloy fuel fission-gas behavior simulations  

SciTech Connect

A mechanistic kinetic rate theory model originally developed for the prediction of fission gas behavior in oxide nuclear fuels under steady-state and transient conditions has been assessed to look at its applicability to model fission gas behavior in U-Pu-Zr metallic alloy fuel. In order to capture and validate the underlying physics for irradiated U-Pu-Zr fuels, the mechanistic model was applied to the simulation of fission gas release, fission gas and fission product induced swelling, and the evolution of the gas bubble size distribution in three different fuel zones: the outer {alpha}-U, the intermediate, and the inner {gamma}-U zones. Due to its special microstructural features, the {alpha}-U zone in U-Pu-Zr fuels is believed to contribute the largest fraction of fission gas release among the different fuel zones. It is shown that with the use of small effective grain sizes, the mechanistic model can predict fission gas release that is consistent with (though slightly lower than) experimentally measured data. These simulation results are comparable to the experimentally measured fission gas release since the mechanism of fission gas transport through the densely distributed laminar porosity in the {alpha}-U zone is analogous to the mechanism of fission gas transport through the interconnected gas bubble porosity utilized in the mechanistic model. Detailed gas bubble size distributions predicted with the mechanistic model in both the intermediate zone and the high temperature {gamma}-U zone of U-Pu-Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreements between the simulation results and experimental measurements, and therefore provide crucial guidelines for the selection of key physical parameters required for modeling these two zones. In addition, the results of parametric studies for several key parameters are presented for both the intermediate zone and the {gamma}-U zone simulations. (authors)

Yun, D.; Rest, J.; Yacout, A. M. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

2012-07-01

44

Fission-gas-release rates from irradiated uranium nitride specimens  

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

45

Data summary report for fission product release test VI-6  

SciTech Connect

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.

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

1994-03-01

46

Fission-product formation in the thermal-neutron-induced fission of odd Cm isotopes  

Microsoft Academic Search

Thermal-neutron-induced fission of 243Cm was studied at the Lohengrin mass separator. The light-mass peak of the fission-yield curve was investigated, and yields of masses from A=72 to A=120 were obtained. Independent-product yields were determined for nuclear charges Z=28 37 . The yield of masses in the superasymmetric region was found to be identical to other fission reactions studied at Lohengrin.

I. Tsekhanovich; N. Varapai; V. Rubchenya; D. Rochman; G. S. Simpson; V. Sokolov; G. Fioni; Ilham Al Mahamid

2004-01-01

47

Prediction of TMI-2 core temperatures from the fission-product release history. Final report  

SciTech Connect

Phenomena associated with fission-gas and fission-product behavior during the accident at TMI-2 are discussed. Calculations performed with the GRASS-SST computer code are used to identify key factors infuencing gas release during TMI-2-type accident conditions. The calculations indicate that fission-product releases estimated from the accident data can be reached by the assumption that rapid fuel temperature increases during the accident led to enhanced bubble mobilities with some type of fuel fracturing. Also, it is concluded from known thermochemical behavior of the fission products that similar magnitudes for the release of Kr, Xe, Cs, and I/sub 2/ indicate that extensive fuel fracturing occurred; however, it is unlikely that temperatures in excess of 2000 K were reached over a large portion of the core.

Rest, J.; Johnson, C.E.

1980-11-01

48

Ceramic Hosts for Fission Products Immobilization  

SciTech Connect

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.

Peter C Kong

2010-07-01

49

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

SciTech Connect

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

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

2013-09-01

50

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

NASA Astrophysics Data System (ADS)

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.

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

51

Yield of the fission products in the fission of 238 Np by thermal neutrons  

Microsoft Academic Search

Obviously, new experimental information on the features of the fission of nuclides with odd Z must provide a clearer picture of the anomalies noted and should explain them. The present paper reports on the results of experimental determinations of the relative and absolute yields of heavy products of the fission effected by thermal neutrons of a shortlived nuclide with odd

A. A. Solonkin; V. F. Teplykh; E. V. Platygina; K. A. Petrzhak; A. V. Mosesov

1988-01-01

52

A novel method for incorporating fission gas elements into solids  

NASA Astrophysics Data System (ADS)

A novel method for the fabrication of test samples for fission gas behavior studies is described. We applied the technique of ion beam assisted deposition (IBAD) as a means to introduce Xe atoms into alumina (Al 2O 3) films. We then investigated the redistribution of Xe atoms and microstructural evolution induced by annealing. Transmission electron microscopy analysis revealed that the microstructure of our Al 2O 3-Xe IBAD films resemble characteristic microstructural features associated with fission gas accumulation in reactor-irradiated nuclear fuels.

Usov, I. O.; Won, J.; Devlin, D. J.; Jiang, Y.-B.; Valdez, J. A.; Sickafus, K. E.

2011-01-01

53

Yields of fission products produced by thermal-neutron fission of 229Th  

NASA Astrophysics Data System (ADS)

Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-?g sample of 229Th. On the basis of measured gamma-ray yields and known nuclear data, yields for cumulative production of 37 fission products were deduced. The absolute overall normalization uncertainty is <8%. The results are compared with fission-product yields previously measured, with generally good agreement. On this basis, and using other measurements for masses not observed in the present experiment, a complete mass distribution for A between 76 and 152 was deduced. The measured A-chain cumulative yields from the present program make up 84% of the total light-mass (A<=115) yield and 77% of the total heavy-mass yield. The data were analyzed to obtain values of most-probable charge (Zp) and charge-dispersion (?) parameters. Based upon insight gained from study of similar data obtained for thermalneutron fission of 235U, we postulate a simple functional dependence ?=?(Zp), and using this dependence obtain values of Zp(A) for 15 mass chains created during fission of 229Th. Values of Zp(A) were estimated for other mass chains based upon results of a recent study of Zp(A). Charge distributions determined using the deduced mass distribution and the deduced sets of Zp(A) and ?(Zp) are in very good agreement with recent measurements, exhibiting a pronounced even-odd effect in elemental yields. These results may be used to predict unmeasured yields for 229Th fission. NUCLEAR REACTIONS Fission 209Th(nth, f), measured fission product gamma-ray yields; deduced fission-product yields and element and mass yields.

Dickens, J. K.; McConnell, J. W.

1983-01-01

54

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

SciTech Connect

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.

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

1987-05-01

55

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

E-print Network

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 ...

Andrews, Nathan Christopher

2012-01-01

56

Fission yields at different fission-product kinetic energies for thermal-neutron-induced fission of 239Pu  

Microsoft Academic Search

At the recoil spectrometer ``Lohengrin'' of the Institut Laue-Langevin in Grenoble, the yields of the light fission products from the thermal-neutron-induced fission of 239Pu were measured as a function of A, Z, the kinetic energy E and the ionic charge states q. The nuclear charge and mass distributions summed over all ionic charge states were determined for different light fissionproduct

C. Schmitt; A. Guessous; J. P. Bocquet; H.-G. Clerc; R. Brissot; D. Engelhardt; H. R. Faust; F. Gnnenwein; M. Mutterer; H. Nifenecker; J. Pannicke; Ch. Ristori; J. P. Theobald

1984-01-01

57

(Fuel, fission product, and graphite technology)  

SciTech Connect

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.

Stansfield, O.M.

1990-07-25

58

Analysis of Fission Products on the AGR-1 Capsule Components  

SciTech Connect

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

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

2013-03-01

59

Fission-product SiC reaction in HTGR fuel  

SciTech Connect

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.

Montgomery, F.

1981-07-13

60

Yields of fission products produced by thermal-neutron fission of 245Cm  

Microsoft Academic Search

Absolute yields have been determined for 105 gamma rays emitted in the decay of 95 fission products representing 54 mass chains created during thermal-neutron fission of 245Cm. These results include 17 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays between 30 sec and 0.3 yr after very short irradiations

J. K. Dickens; J. W. McConnell

1981-01-01

61

Yields of fission products produced by thermal-neutron fission of 229Th  

Microsoft Academic Search

Absolute yields have been determined for 47 gamma rays emitted in the decay of 37 fission products representing 25 mass chains created during thermal-neutron fission of 229Th. Using a Ge(Li) detector, spectra were obtained of gamma rays emitted between 15 min and 0.4 yr after very short irradiations by thermal neutrons of a 15-mug sample of 229Th. On the basis

J. K. Dickens; J. W. McConnell

1983-01-01

62

Yields of fission products produced by thermal-neutron fission of ²⁴⁹Cf  

Microsoft Academic Search

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of ²⁴⁹Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 ..mu..g sample of ²⁴⁹Cf between 45

J. K. Dickens; J. W. McConnell

1981-01-01

63

Yields of fission products produced by thermal-neutron fission of 249Cf  

Microsoft Academic Search

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 mug sample of 249Cf between 45

J. K. Dickens; J. W. McConnell

1981-01-01

64

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

SciTech Connect

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

J M Harp; P D Demkowicz; S A Ploger

2012-10-01

65

Fission-product formation in the thermal-neutron-induced fission of odd Cm isotopes  

NASA Astrophysics Data System (ADS)

Thermal-neutron-induced fission of 243Cm was studied at the Lohengrin mass separator. The light-mass peak of the fission-yield curve was investigated, and yields of masses from A=72 to A=120 were obtained. Independent-product yields were determined for nuclear charges Z=28 37 . The yield of masses in the superasymmetric region was found to be identical to other fission reactions studied at Lohengrin. The multimodal approach to fission and the macroscopic-microscopic method for the calculation of charge-distribution parameters in isobaric chains were used to analyze experimental results from the fission of 243Cm and 245Cm . A systematics on fission modes was derived from the analysis and extended to the 247Cm case. The weight of the 132Sn mode was found to decrease in 243Cm , relative to the 245Cm nucleus. A prediction of the 78Ni yield in the fission of Cm isotopes was made. The feasibility of the study of 78Ni at Lohengrin has been demonstrated.

Tsekhanovich, I.; Varapai, N.; Rubchenya, V.; Rochman, D.; Simpson, G. S.; Sokolov, V.; Fioni, G.; Mahamid, Ilham Al

2004-10-01

66

Thermodynamics of fission products in UO2+-x  

SciTech Connect

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.

Nerikar, Pankaj V [Los Alamos National Laboratory

2009-01-01

67

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

SciTech Connect

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)

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

2012-07-01

68

Fission yields at different fission-product kinetic energies for thermal-neutron-induced fission of 239Pu  

NASA Astrophysics Data System (ADS)

At the recoil spectrometer "Lohengrin" of the Institut Laue-Langevin in Grenoble, the yields of the light fission products from the thermal-neutron-induced fission of 239Pu were measured as a function of A, Z, the kinetic energy E and the ionic charge states q. The nuclear charge and mass distributions summed over all ionic charge states were determined for different light fissionproduct kinetic energies between 93 and 112 MeV. The proton odd-even effect which was measured to be (11.6 0.6)% causes considerable fine structure in the yields. The average kinetic energy of even- Z elements in the light fission-product group is 0.3 0.1 MeV larger than for odd- Z elements. The neutron odd-even effect is (6.5 0.7)%. The comparison with previously published data 1) for thermal-neutron-induced fission of 235U reveals a correlation between the proton odd-even effect in the yield and in the kinetic energy of the elements. The dependence of the proton odd-even effect on the fragmentation is very similar for 235U and 239Pu when it is considered as a function of the nuclear charge of the heavy fission products. The isobaric variances ?z2. for thermal-neutron fission of 235U and 239Pu coincide at all kinetic energies if the influence of the proton odd-even effect is averaged out. This supports the hypothesis that the magnitude of ?z2 is determined only by quantum-mechanical zero-point fluctuations. The influence of the spherical shells Z = 50 and N = 82 on the fragmentation is discussed.

Schmitt, C.; Guessous, A.; Bocquet, J. P.; Clerc, H.-G.; Brissot, R.; Engelhardt, D.; Faust, H. R.; Gnnenwein, F.; Mutterer, M.; Nifenecker, H.; Pannicke, J.; Ristori, CH.; Theobald, J. P.

1984-11-01

69

Data summary report for fission product release test HI6  

Microsoft Academic Search

The sixth in a series of high-temperature fission product release tests was conducted for 1 min at 1950°C in a steam-helium atmosphere. The 15.2-cm-long test specimen was a section of fuel rod which was irradiated to 40.3 MWd\\/kg in the Monticello BWR. Based on fission product inventories, analyses of test components by gamma spectrometry and neutron activation showed total releases

M. F. Osborne; R. A. Lorenz; K. S. Norwood; J. R. Travis; C. S. Webster; Jack Lee Collins

1985-01-01

70

ESTIMATION OF THE FISSION PRODUCTS, ACTINIDES AND TRITIUM OF HTR-10  

Microsoft Academic Search

Given the evolution of High-Temperature Gas-cooled Reactor (HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for

HYEDONG JEONG; SOON HEUNG CHANG

71

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

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.

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

2007-10-01

72

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

NASA Astrophysics Data System (ADS)

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.

Padgett, Stephen; Wang, Tzu-Fang

2015-01-01

73

Studies of short-lived products of spallation fission reactions at TRIUMF  

E-print Network

The gas-jet recoil transport technique has been used to transport products from spallation and fission reactions from a target chamber to a shielded location for nuclear spectroscopic studies. These involve X- beta - gamma coincidence measurements and (shortly) time- of-flight mass spectroscopy. It has been deduced that the proton beam at present intensities has no appreciable effect on the ability of ethylene and other cluster-producing gases to transport radioactivity. Preliminary results will be presented for shortlived fission products from uranium, and for spallation products of iodine and argon. The latter were obtained from the bombardment of gas and aerosol targets mixed with the transporting gas in the target chamber, which appears to be a generally useful technique.

Bischoff, G; D'Auria, J M; Dautet, H; Lee, J K P; Pate, B D; Wiesehahn, W

1976-01-01

74

Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions  

SciTech Connect

The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted.

Rest, J.

1985-10-01

75

Yields of fission products produced by thermal-neutron fission of 245Cm  

NASA Astrophysics Data System (ADS)

Absolute yields have been determined for 105 gamma rays emitted in the decay of 95 fission products representing 54 mass chains created during thermal-neutron fission of 245Cm. These results include 17 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays between 30 sec and 0.3 yr after very short irradiations of thermal neutrons on a 1 ?g sample of 245Cm. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 84 and 156. The absolute overall normalization uncertainty is <8%. The measured A-chain cumulative yields make up 81% of the total light mass (A<=121) yield and 92% of the total heavy mass yield. The results are compared with fission-product yields previously measured with generally good agreement. The mass-yield data have been compared with those for thermal-neutron fission of 239Pu and for 252Cf(s.f.); the influences of the closed shells Z=50, N=82 are not as marked as for thermal-neutron fission of 239Pu but much more apparent than for 252Cf(s.f.). Information on the charge distribution along several isobaric mass chains was obtained by determining fractional yields for 12 fission products. The charge distribution width parameter, based upon data for the heavy masses, A=128 to 140, is independent of mass to within the uncertainties of the measurements. Gamma-ray assignments were made for decay of short-lived fission products for which absolute gamma-ray transition probabilities are either not known or in doubt. Absolute gamma-ray transition probabilities were determined as (51 +/- 8)% for the 374-keV gamma ray from decay of 110Rh, (35 +/- 7)% for the 1096-keV gamma ray from decay of 133Sb, and (21.2 +/- 1.2)% for the 255-keV gamma ray from decay of 142Ba. RADIOACTIVITY, FISSION 245Cm(n,f) En=thermal; measured ?(E?,T12) deduced mass, charge yields.

Dickens, J. K.; McConnell, J. W.

1981-01-01

76

The behavior of fission products during nuclear rocket reactor tests  

SciTech Connect

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.

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

1991-01-01

77

The behavior of fission products during nuclear rocket reactor tests  

NASA Astrophysics Data System (ADS)

The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955 to 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 a 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.

Bokor, Peter C.; Kirk, William L.; Bohl, Richard J.

78

Evaluation and compilation of fission product yields 1993  

SciTech Connect

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.

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

1995-12-31

79

Effects of time and other variables on fission product release rates  

SciTech Connect

The releases of krypton and cesium from highly irradiated LWR fuel have been examined in detail. The main interest has been the effect of time on the rate of release and the effects of heatup and cooldown cycles. The minute-by-minute release rates for fission product /sup 85/Kr from commercial fuel irradiated in the H.B. Robinson PWR are shown. The release rate, fraction per minute, is calculated in the same manner as release rates given in NUREG-0772; the fission gas, cesium, and iodine release rate curve from that report is also shown.

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

1986-01-01

80

Fission product removal from molten salt using zeolite  

SciTech Connect

Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission product cations from the molten salt. Breakthrough of cesium and the alkaline earths occurred more rapidly than was observed for the rare earths. The effluent composition as a function of time is presented, as well as results for the distribution of fission products along the length of the column. Effects of temperature and salt flow rate are also discussed.

Pereira, C.; Babcock, B.D.

1996-10-01

81

Comparison of Fission Product Yields and Their Impact  

SciTech Connect

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.

S. Harrison

2006-02-01

82

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

SciTech Connect

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.

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

83

Yields of fission products produced by thermal-neutron fission of 249Cf  

NASA Astrophysics Data System (ADS)

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 ?g sample of 249Cf between 45 s and 0.4 yr after very short irradiations of the 249Cf by thermal neutrons. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 89 and 156. The absolute overall normalization uncertainty is ~8%. The measured A-chain cumulative yields make up 77% of the total light mass (A<=123) yield and 79% of the total heavy mass yield. The results are compared with fission-product yields previously measured, with generally good agreement. Information on the charge distribution along several isobaric mass chains was obtained by determining fractional yields for 11 fission products and combining these results with other measurements. The charge distribution width parameter for the heavy masses A=128 to 140 is independent of mass to within the uncertainties of the measurements. For the light masses A=89 to 112 the charge distribution parameter is also independent of mass but is smaller than for the heavy masses. Total chain yields are in fair agreement with the current evaluation for 249Cf. [RADIOACTIVITY, FISSION 249Cf(n,f) En=thermal measured ?(E?,T12) deduced mass, charge yields.

Dickens, J. K.; McConnell, J. W.

1981-07-01

84

An efficient model for the analysis of fission gas release  

NASA Astrophysics Data System (ADS)

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.

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

2002-04-01

85

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

NASA Astrophysics Data System (ADS)

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.

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

86

Fission-product energy release for times following thermal-neutron fission of U between 2 and 14000 seconds  

Microsoft Academic Search

Fission-product decay energy-releases rates were measured for thermal-neutron fission of U. Samples of mass 1 to 10 ..mu..g were irradiated for 1 to 100 sec by use of the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were counted for times-after-fission between 2 and 14,000 seconds. The data were obtained for beta and

J. K. Dickens; J. F. Emery; T. A. Love; J. W. McConnell; K. J. Northcutt; R. W. Peelle; H. Weaver

1977-01-01

87

Fission product release from highly irradiated LWR fuel  

SciTech Connect

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.

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

1980-02-01

88

New Fission-Product Waste Forms: Development and Characterization  

Microsoft Academic Search

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

Alexandra Navrotsky

2010-01-01

89

Data summary report for fission product release test VI-5  

SciTech Connect

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.

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

1991-10-01

90

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

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 2,000C 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.

Dawn Scates

2010-10-01

91

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

SciTech Connect

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.

Blaise Collin

2014-09-01

92

Design of pellet surface grooves for fission gas plenum  

SciTech Connect

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

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

1986-01-01

93

Nuclear decay studies of rare-earth fission-product nuclides using fast radiochemical separation techniques  

Microsoft Academic Search

A facility for the rapid radiochemical separation of individual rare-earth fission product nuclides from mixed fission products has been developed at the Idaho National Engineering Laboratory (INEL). This facility, called the INEL ESOL (elemental separation on-line) facility, includes an electroplated spontaneously fissioning252Cf source, a He jet transport system to deliver short half-life fission products from the252Cf hot cell to the

J. D. Baker; D. H. Meikrantz; R. C. GREENWOOD

1990-01-01

94

Nuclear decay studies of fission-product nuclides using an on-line mass separation technique  

Microsoft Academic Search

An isotope-separator-on-line (ISOL) system has been developed at the Idaho National Engineering Laboratory to enable a wide variety of nuclear decay studies to be made for fission-product radionuclides. The system is unique in that it utilizes the spontaneous fission source,252Cf, as the source of fission-product radioactivity. Fission products are transported to the ion source of the mass separator by the

R. A. Anderl; R. C. Greenwood

1990-01-01

95

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

NASA Astrophysics Data System (ADS)

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.

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

96

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

SciTech Connect

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.

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

97

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

SciTech Connect

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.

Christopher Matthews; Daniel Schwen; Andrew C. Klein

2014-12-01

98

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

NASA Astrophysics Data System (ADS)

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.

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

2015-02-01

99

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

SciTech Connect

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)

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

1983-01-01

100

Chemistry of fission product iodine in light-water reactors  

SciTech Connect

Control of fission product iodine in LWRs have been based on the assumption that the bulk of the iodine is uncombined chemically. Recent studies, however, have indicated that fission product iodine released from LWR fuel rods with defected cladding is combined chemically with electropositive species. Thermodynamic studies indicate that the most likely form of the iodine is cesium iodide. In most cases, the chemical form of the radioiodine in the vapor phase is determined by reactions in the aqueous phase. Studies of aqueous iodine chemistry indicate that the predominant dissolved species are iodide and iodate ions in relative concentrations determined by the redox conditions involved. Moreover, volatile species over the aqueous system can be maintained at very low levels. Work performed to date suggests that the dominant chemical form of radioiodine to be controlled in a reactor containment building (excluding aerosols) is methyl iodide, albeit at low concentration.

Malinauskas, A.P.; Bell, J.T.; Campbell, D.O.; Lorenz, R.A.

1981-01-01

101

Fission-product release from irradiated LWR fuel  

SciTech Connect

An experimental investigation of fission product release from commercial LWR fuel under accident conditions is being conducted at Oak Ridge National Laboratory (ORNL). This work, which is sponsored by the US Nuclear Regulatory Commission (NRC), is an extension of earlier experiments up to 1600/sup 0/C and is designed to obtain the experimental data needed to reliably assess the consequences of accidents for fuel temperatures up to melting. The objectives of this program are (1) to determine fission product release rates from fully-irradiated commercial LWR fuel in high-temperature steam; (2) to collect and characterize the aerosol released; (3) to identify the chemical forms of the released material; (4) to correlate the results with related experimental data and develop a consistent source term model; and (5) to aid in the interpretation of tests using simulated LWR fuel.

Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

1982-01-01

102

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

DOEpatents

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.

Herrmann, Steven Douglas

2014-05-27

103

Fission track astrology of three Apollo 14 gas-rich breccias  

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

104

Fission product ion exchange between zeolite and a molten salt  

NASA Astrophysics Data System (ADS)

The electrometallurgical treatment of spent nuclear fuel (SNF) has been developed at Argonne National Laboratory (ANL) and has been demonstrated through processing the sodium-bonded SNF from the Experimental Breeder Reactor-II in Idaho. In this process, components of the SNF, including U and species more chemically active than U, are oxidized into a bath of lithium-potassium chloride (LiCl-KCl) eutectic molten salt. Uranium is removed from the salt solution by electrochemical reduction. The noble metals and inactive fission products from the SNF remain as solids and are melted into a metal waste form after removal from the molten salt bath. The remaining salt solution contains most of the fission products and transuranic elements from the SNF. One technique that has been identified for removing these fission products and extending the usable life of the molten salt is ion exchange with zeolite A. A model has been developed and tested for its ability to describe the ion exchange of fission product species between zeolite A and a molten salt bath used for pyroprocessing of spent nuclear fuel. The model assumes (1) a system at equilibrium, (2) immobilization of species from the process salt solution via both ion exchange and occlusion in the zeolite cage structure, and (3) chemical independence of the process salt species. The first assumption simplifies the description of this physical system by eliminating the complications of including time-dependent variables. An equilibrium state between species concentrations in the two exchange phases is a common basis for ion exchange models found in the literature. Assumption two is non-simplifying with respect to the mathematical expression of the model. Two Langmuir-like fractional terms (one for each mode of immobilization) compose each equation describing each salt species. The third assumption offers great simplification over more traditional ion exchange modeling, in which interaction of solvent species with each other is considered. (Abstract shortened by UMI.)

Gougar, Mary Lou D.

105

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

106

Long-lived fission product transmutation in nuclear reactors  

SciTech Connect

One of the main directions in the management of high-level radioactive wastes is the development of specialized reactors for transmutation with maximum support coefficients for the existing power reactor. The developments have shown that it is more expetitious to design the reactor for actinide transmutation and for fission products separately. For the above purposes, the FBR type fast neutron reactor and FMF type fast reactor with melted fuel were considered.

Ganev, I.K.; Lopatkin, A.V.; Naumov, V.V.; Reshetov, V.A.

1993-12-31

107

Measurement of Isomeric Yield Ratios of Fission Products with the Jyfltrap  

NASA Astrophysics Data System (ADS)

The fission system at the scission configuration is characterized by the angular momentum of the initial fission fragments. The angular momentum of the primary fragments can be deduced from the independent isomeric yield ratio of fission products. Usually such ratios are measured by spectroscopic methods. Nevertheless, completely different method, utilizing capabilites of the double Penning-trap mass spectrometer JYFLTRAP, can be used to measure the independent isomeric yield ratio of fission products.

Gorelov, D.; Penttil, H.; Igisol Group,; Lantz, M.; Mattera, A.; Pomp, S.

2014-09-01

108

Fission product release from fuel under LWR accident conditions  

SciTech Connect

Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 2000/sup 0/C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gamma spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species.

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

1983-01-01

109

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

Microsoft Academic Search

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

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

1983-01-01

110

Volatile fission-product source term evaluation using the fastgrass comptuer code. [PWR; BWR  

SciTech Connect

As the noble gases play a major role in establishing the interconnection of escape routes from the interior to the exterior of nuclear reactor fuel, a realistic description of the release of volatile fission products (VFPs) must a priori include a realistic description of fission-gas release and swelling. The steady-state and transient gas release and swelling subroutine, FASTGRASS, has been modified to include a mechanistic description of behavior of VFPs (I, Cs, CsI, Cs/sub 2/MoO/sub 4/, and Cs/sub 2/UO/sub 4/). Phenomena modeled are the chemical reactions between the VFPs, VFP migration through the fuel, and VFP interaction with the noble gases. This paper will describe calculations performed with FASTGRASS to describe the release of noble gases, I, Cs, and CsI from LWR fuel during steady-state and power-ramping conditions. Key issues that are addressed in the analysis are the effects of (a) VFP chemistry, (b) various assumptions concerning mechanisms of VFP migration through solid UO/sub 2/, (c) fission-gas behavior, and (d) accident scenario on the chemical form of iodine and the rate of iodine release from water-reactor fuel.

Rest, J.

1982-08-01

111

Fusion-Fission Hybrid for Fissile Fuel Production without Processing  

SciTech Connect

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.

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

2012-01-02

112

Grain Boundary Percolation Modeling of Fission Gas Release in Oxide Fuels  

SciTech Connect

We present a new approach to fission gas release modeling in oxide fuels based on grain boundary network percolation. The method accounts for variability in the bubble growth and coalescence rates on individual grain boundaries, and the resulting effect on macroscopic fission gas release. Two-dimensional representa- tions of fuel pellet microstructures are considered, and the resulting gas release rates are compared with traditional two-stage Booth models, which do not account for long-range percolation on grain boundary net- works. The results show that the requirement of percolation of saturated grain boundaries can considerably reduce the total gas release rates, particularly when gas resolution is considered.

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

2012-05-01

113

Release of volatile fission products from uranium dioxide  

SciTech Connect

Post-irradiation anneal experiments have been used to determine the release of iodine and tellurium from lightly irradiated UO/sub 2/ samples maintained at stoichiometry. The applicability of the equivalent-sphere model of diffusion to release of fission gases has been tested. Diffusion coefficients and activation energies have been evaluated. The diffusion coefficient of /sup 132/Te at 1400/sup 0/C was found to be of an order-of-magnitude larger than that of /sup 131/I. This result may be of importance for an understanding of the pellet-cladding interaction and for a better evaluation of the source term for fission-product release under accident conditions. Qualitatively, the influence of the stoichiometry on the release of /sup 133/Xe, /sup 131/I, and /sup 132/Te has been established.

Bayen, D.

1983-03-01

114

THE EXPERIMENTAL EVALUATION OF FISSION PRODUCT MOBILITY. I. EQUIPMENT AND CALIBRATION STUDIES  

Microsoft Academic Search

An apparatus is described for use in fission gas mobility studies using ; post-irradiation annealing techniques. Basically the apparatus comprises a ; resistance furnace, temperature controlling and recording equipment, gamma ray ; scintillation spectrometer, and a ceramic sample tube, containing specimen, ; thermocouple, and pressure gauge. Calibration data are presented for the ; measurement of fission gas Xe¹³³, taking into

Moses

1963-01-01

115

Radiolytic gas production from concrete containing Savannah River Plant waste  

Microsoft Academic Search

To determine the extent of gas production from radiolysis of concrete containing radioactive Savannah River Plant waste, samples of concrete and simulated waste were irradiated by Co gamma rays and Cm alpha particles. Gamma radiolysis simulated radiolysis by beta particles from fission products in the waste. Alpha radiolysis indicated the effect of alpha particles from transuranic isotopes in the waste.

Bibler

1978-01-01

116

Venting of fission products and shielding in thermionic nuclear reactor systems  

NASA Technical Reports Server (NTRS)

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.

Salmi, E. W.

1972-01-01

117

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

SciTech Connect

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.

Jason M. Harp; Paul A. Demkowicz

2014-10-01

118

Yield, kinetic energy, pairing effect, and shell effect of light fission products for thermal-neutron fission of uranium 233  

Microsoft Academic Search

The mass spectrometer HIAWATHA in conjunction with a thin-windowed gridded ionization chamber was used to determine the yields of light fission products for thermal-neutron fission of U-233. HIAWATHA has an energy resolution of 0.3%, a mass resolution of 0.5 amu, and an atomic-number resolving power of 38. The yields were measured as a function of mass number, atomic number, and

1983-01-01

119

Nuclide yields of light fission products from thermal-neutron induced fission of 233U at different kinetic energies  

Microsoft Academic Search

The yields of light fission products from thermal-neutron induced fission of 233U are measured as a function of their mass A, their nuclear charge Z, their kinetic energy E and their ionic charge state q at the recoil spectrometer Lohengrin of the Institut Laue-Langevin in Grenoble. The mass yields are determined by intercepting the fragments with an ionization chamber of

U. Quade; K. Rudolph; S. Skorka; P. Armbruster; H.-G. Clerc; W. Lang; M. Mutterer; C. Schmitt; J. P. Theobald; F. Gnnenwein; J. Pannicke; H. Schrader; G. Siegert; D. Engelhardt

1988-01-01

120

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

SciTech Connect

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.

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

121

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

SciTech Connect

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.

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

1983-10-01

122

Engineering Report on the Fission Gas Getter Concept  

SciTech Connect

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).

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

2012-11-01

123

Measurement and characterization of fission products released from LWR fuel  

SciTech Connect

Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400/sup 0/C to >50% at 2000/sup 0/C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag.

Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

1984-01-01

124

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

NASA Astrophysics Data System (ADS)

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

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

2011-09-01

125

Removal of Fission Products and Their Complexing Agents from Degraded Solvent by Ion Exchange Method  

Microsoft Academic Search

Removal of fission products and their complexing agents from chemically degraded TBP\\/kerosene and particularly from the kerosene diluent, was studied. With the use of column technique with mixed anion and cation exchange resins, more than 80% of the fission products and their complexing agents were removed. The carboxylic acids, which constitute one of the degradation products bringing about decrease of

Ken OHWADA

1967-01-01

126

Studies of fission product movement in tuffaceous media  

SciTech Connect

For approximately 25 years the United States has conducted underground nuclear tests at a site in the state of Nevada. These tests have left a variety of fission products at depths of 100 to 1000 meters below the land surface. The geologic media here consist primarily of tuffs and rhyolites. More than 150 tests were conducted at or below the water table. We are studying locations of past tests to determine whether residual fission products move through the underground environment and, if so, by what mechanisms. Our research involves consideration of leaching, sorption, hydraulic dispersion, fracture flow and colloid transport. The data we obtain are relevant to groundwater contamination and nuclear waste storage issues. In this paper we present information obtained from our research at several different locations within the study site. Specifically, we describe the movement of radionuclides including tritium, {sup 85}Kr, {sup 90}Sr, {sup 106}Ru, {sup 125}Sb, and {sup 137}Cs in situations were groundwater was moving and in which it was relatively static. 15 refs., 2 figs.

Thompson, J.L.

1991-09-01

127

Assessment of selected fission products in the Savannah River Site environment  

SciTech Connect

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.

Carlton, W.H.; Denham, M.

1997-04-01

128

Measurement of cumulative and independent yields of fission products from thermal-neutron fission of ²⁴²\\/sup m\\/ Am  

Microsoft Academic Search

The mass and charge distributions in an unseparated mix of fission product nuclei from thermal-neutron fission of \\/sup 242m\\/Am were studied through semiconductor gamma-ray spectrometry. Samples of the fissionable material under study were irradiated in a vertical irradiation tube of the MIFI IRT research reactor. Following irradiation, measurements were made on aperture-calibrated semiconductor detectors. For broader identification of fission fragment

A. N. Gudkov; V. M. Zhivun; V. V. Kovalenko; A. B. Koldobskii; V. M. Kolobashkin; C. V. Krivasheev; N. S. Piven; E. V. Semenova; V. A. Khristoforov

1985-01-01

129

Yields of rare earth fission products in the spontaneous fission of californium-252  

Microsoft Academic Search

Cumulative fission yields of rare earth isotopes have been determined in the spontaneous fission of252Cf by fast radiochemical separation and gamma-ray spectrometry. The determined yield values are compared with the available literature data. The yield values for147Nd,151Nd and151Pm differ from the reported values. The yield for145Ce is determined for the first time.

B. S. Tomar; H. Naik; A. Ramaswamy; Satya Prakash

1985-01-01

130

CJ) Gas Products GAS EQUIPMENT TECHNOLOGY GROUP  

E-print Network

UNITS: TUBE NUMBER; · SERIAL NUMBER: FLOAT MATERIAL; CERT FILE #: SCCM WATER E501 TYPICAL GLASS 'E500WGC£±J) Gas Products GAS EQUIPMENT TECHNOLOGY GROUP UN\\TS: TUBE NUMBER: SERIAL NUMBER: FLOAT MATERIAL TYPICAL GLASS E500G SCALE READING DATE: STD CONDITIONS: GAS TEMPERATURE; PRESSURE IN TUBE: 2/27/97 1 ATMOS

Kleinfeld, David

131

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

SciTech Connect

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.

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

1982-08-01

132

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

DOEpatents

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.

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

1987-10-26

133

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

DOEpatents

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.

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

1989-01-01

134

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

SciTech Connect

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.

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

2009-11-01

135

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

SciTech Connect

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

Malinauskas, A.P.

1985-01-01

136

Assessment of fission product yields data needs in nuclear reactor applications  

SciTech Connect

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)

Kern, K.; Becker, M.; Broeders, C. [Institut fuer Neutronenphysik und Reaktortechnik, KIT Campus Nord, Hermann-von-Helmholtz-Platz 1, 76344 Leopoldshafen (Germany)

2012-07-01

137

Decontamination of actinides and fission products from stainless steel surfaces  

SciTech Connect

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

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

1996-04-01

138

Use of fission product implantation in nuclear waste management  

NASA Astrophysics Data System (ADS)

During the reactor processing, fission products among which iodine are implanted by recoil inside the zircalloy cladding tube: most of them being distributed in the first 2 ?m. At the same time oxidation of the cladding tube occurs, hence in the waste storage phase zirconia will act as a migration barrier. In order to determine diffusion data, stable and radioactive iodine atoms were introduce in zirconium oxidized samples by mean of ion implantation. Iodine thermal-release was measured either by Rutherford Backscattering Spectroscopy or ? spectroscopy. Two depths range were studied, the subsurface (<30 nm) and one micrometer mean range. The analysis of the iodine release data so obtained allows to determine diffusion coefficients and activation energies.

Brossard, F.; Carlot, G.; Chevarier, A.; Chevarier, N.; Crusset, D.; Duclot, J. C.; Faust, H.; Gaillard, C.; Millard-Pinard, N.; Moncoffre, N.

1998-10-01

139

Data summary report for fission product release test HI-6  

SciTech Connect

The sixth in a series of high-temperature fission product release tests was conducted for 1 min at 1950/sup 0/C in a steam-helium atmosphere. The 15.2-cm-long test specimen was a section of fuel rod which was irradiated to 40.3 MWd/kg in the Monticello BWR. Based on fission product inventories, analyses of test components by gamma spectrometry and neutron activation showed total releases of 29.6% for /sup 85/Kr, 33.1% for /sup 137/Cs, 24.7% for /sup 129/I, 6.0% for /sup 110m/Ag, and 0.06% for /sup 125/Sb. Rubidium and bromine were also detected in the collection system by spark-source mass spectrometry. The fractions released of these chemical analogs of cesium and iodine should have been similar to those obtained for cesium and iodine. Cesium reacted with the oxidized surface in the stainless steel thermal gradient tube; steam oxidation occurred at temperatures greater than or equal to600/sup 0/C. Cesium that reacted at approx.800/sup 0/C was considerably more difficult to remove by leaching with simulated ''LWR coolant'' at 53/sup 0/C than cesium that reacted or otherwise deposited at approx.600/sup 0/C. The thermal gradient tube deposition profile indicated that iodine probably deposited as cesium iodide. A comparison was made of Cs, I, and Kr release rate coefficients obtained in the HI and HT test series with NUREG-0772 values. The coefficients obtained in the HI tests were factors of 19 and 3.2 times lower than the HT tests and NUREG-0772 values, respectively. 23 refs., 19 figs., 18 tabs.

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

1985-09-01

140

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

E-print Network

requires recycling of useful materials from spent nuclear fuel and discarding of non-usable streams of the spent fuel, which are predominantly the fission products. The fission products represent the near-term concerns associated with final geological...

Alajo, Ayodeji Babatunde

2011-08-08

141

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

Microsoft Academic Search

Within quantum fission theory, angular distributions of products originating from the ternary fission of nuclei that is induced\\u000a by polarized cold and thermal neutrons are investigated on the basis of a nonevaporative mechanism of third-particle emission\\u000a and a consistent description of fission-channel coupling. It is shown that the inclusion of Coriolis interaction both in the\\u000a region of the discrete and

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

2008-01-01

142

Linking photochemistry in the gas and solution phase: S-H bond fission in p-methylthiophenol following UV photoexcitation.  

PubMed

Gas-phase H (Rydberg) atom photofragment translational spectroscopy and solution-phase femtosecond-pump dispersed-probe transient absorption techniques are applied to explore the excited state dynamics of p-methylthiophenol connecting the short time reactive dynamics in the two phases. The molecule is excited at a range of UV wavelengths from 286 to 193 nm. The experiments clearly demonstrate that photoexcitation results in S-H bond fission--both in the gas phase and in ethanol solution-and that the resulting p-methythiophenoxyl radical fragments are formed with significant vibrational excitation. In the gas phase, the recoil anisotropy of the H atom and the vibrational energy disposal in the p-MePhS radical products formed at the longer excitation wavelengths reveal the operation of two excited state dissociation mechanisms. The prompt excited state dissociation motif appears to map into the condensed phase also. In both phases, radicals are produced in both their ground and first excited electronic states; characteristic signatures for both sets of radical products are already apparent in the condensed phase studies after 50 fs. No evidence is seen for either solute ionisation or proton coupled electron transfer--two alternate mechanisms that have been proposed for similar heteroaromatics in solution. Therefore, at least for prompt S-H bond fissions, the direct observation of the dissociation process in solution confirms that the gas phase photofragmentation studies indeed provide important insights into the early time dynamics that transfer to the condensed phase. PMID:22457960

Oliver, Thomas A A; Zhang, Yuyuan; Ashfold, Michael N R; Bradforth, Stephen E

2011-01-01

143

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

SciTech Connect

Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show an overall over-prediction of the fractional release of these fission products, which is largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. Correction factors to these diffusivities were assessed for silver and cesium in order to enable a better match between the modeling predictions and the safety testing results. In the case of strontium, correction factors could not be assessed because potential release during the safety tests could not be distinguished from matrix content released during irradiation. In the case of krypton, all the coating layers are partly retentive and the available data did not allow to determine their respective retention powers, hence preventing to derive any correction factors.

Blaise P. Collin

2001-10-01

144

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

NASA Astrophysics Data System (ADS)

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.

Barber, Duncan Henry

145

A systematics of fission product mass yields with 5 gaussian functions  

E-print Network

A systematics of fission product mass yields is proposed. The systematics is based on Moriyama-Ohnishi systematics developed about 30 years ago. The parameter set of the systematics is newly determined by examining measured data taken after Moriyama-Ohnishi systematics was released. The systematics using the newly determined parameter set is employed to calculate mass distributions of various kinds of fission and compare them with measured data. The comparison shows rather good agreement between them from spontaneous fission to high energy particle induced fission.

Katakura, J I

2003-01-01

146

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

NASA Technical Reports Server (NTRS)

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.

1973-01-01

147

Fusion-Fission Hybrid for Fissile Fuel Production without Processing  

Microsoft Academic Search

Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of Th and fission of U in situ without reprocessing or 'closed' cycles based on irradiation of Th followed by reprocessing, and recycling of U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile

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

2012-01-01

148

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

SciTech Connect

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)

Ajlouni, Abdul-Wali M.S. [Ministry of Energy and Mineral Resources, Amman 11814 (Jordan)

2006-07-01

149

Background and Derivation of ANS-5.4 Standard Fission Product Release Model  

SciTech Connect

This background report describes the technical basis for the newly proposed American Nuclear Society (ANS) 5.4 standard, Methods for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuels. The proposed ANS 5.4 standard provides a methodology for determining the radioactive fission product releases from the fuel for use in assessing radiological consequences of postulated accidents that do not involve abrupt power transients. When coupled with isotopic yields, this method establishes the 'gap activity,' which is the inventory of volatile fission products that are released from the fuel rod if the cladding are breached.

Beyer, Carl E.; Turnbull, Andrew J.

2010-01-29

150

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

SciTech Connect

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

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

2012-02-01

151

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

E-print Network

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.

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

152

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

SciTech Connect

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

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

2013-03-01

153

Fission-product energy release for times following thermal-neutron fission of ²⁴¹Pu between 2 and 14,000 seconds  

Microsoft Academic Search

Fission-product decay energy-release rates were measured for thermal-neutron fission of ²⁴¹Pu. Samples of mass 1 and 5 ..mu..g were irradiated for 1 to 50 s by use of the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emission spectra were recorded for times-after-fission between 2 and 14,000 s. The data were obtained for beta

J. K. Dickens; J. F. Emery; T. A. Love; J. W. McConnell; K. J. Northcutt; R. W. Peelle; H. Weaver

1978-01-01

154

Fission products from the damaged Fukushima reactor observed in Hungary.  

PubMed

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

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

2014-01-01

155

Fission product release from Triso-coated UO 2 particles at 1940 to 2320C  

NASA Astrophysics Data System (ADS)

The fission product release from TRISO-coated UO 2 particles was measured by post-activation heating at 1940 to 2320C for use in a safety analysis. The results are analyzed mathematically with effective diffusion coefficients in each medium. 103Ru, 99Mo and 95Nb are released at 1940 to 2320C and have high effective diffusion coefficients. Although 140Ba and 137Cs are retained in TRISO-coated particles at 2050C, they are released rapidly at 2320C. This is attributed to the transition of beta to alpha SiC at 2320C. 141Ce, 140La and 95Zr are released little if any at 2320C. Rare gas nuclides, iodine and tellurium seem to be retained in coated particles at this high temperature.

Kurata, Yuji; Ikawa, Katsuichi; Iwamoto, Kazumi

1981-05-01

156

Behavior of Fission Products in YSZ-Based Inert Matrix Fuel  

SciTech Connect

Pu disposal has led to increased interest in the possibility of ''burning'' actinides in inert-matrix fuels. Yttria-stabilized cubic zirconia (YSZ) is a promising candidate material. The effects of fission product incorporation on the microstructure of YSZ (with 9.5 mol% of yttria) were investigated by ion implantation (using 70- to 440-keV Cs{sup +}, Sr{sup +}, and I{sup +} ions) and transmission electron microscopy (TEM) in order to evaluate the material's performance as both an inert fuel matrix and a nuclear waste form. It was found that incorporation of an excess amount of cesium (>8 at.%) at room temperature causes amorphization of the cubic zirconia structure, which may lead to a higher leaching rate in the waste repository. On the other hand, iodine and strontium precipitate out in gas bubbles or secondary phases, respectively, at elevated temperature, leading to a lower release rate of the radionuclides.

Wang, L.M.; Zhu, S.; Ewing, R.C.

2001-06-17

157

RETENTION CAPACITIES OF ZIRCONIA AND APATITES TOWARDS IODINE AND TECHNETIUM FISSION PRODUCTS  

E-print Network

1 RETENTION CAPACITIES OF ZIRCONIA AND APATITES TOWARDS IODINE AND TECHNETIUM FISSION PRODUCTS N of technetium. Two materials were more particularly studied: First, apatites whose general formula is Ca10(PO4

Paris-Sud XI, Universit de

158

Superasymmetric Fission  

NASA Astrophysics Data System (ADS)

The existence of superasymmetric fission mode connected with Z = 28 and N = 50 nuclear shells is analysed in the framework of the scission point model. Calculations of PES near the scission point had shown that the 78Ni fission mode would be manifested in fission of neutron-rich compound nuclei. In the case of fission of superheavy nucleus the superasymmetric fission mode is enhanced by influence of the Z = 82 and N = 126 nuclear shells in heavy fragment. Enhancement of highly asymmetric mass and charge division in the proton and neutron fission of 238U at intermediate energy in comparison with thermal neutron induced fission was described by the model developed for calculating the product yields with inclusion of superasymmetric fission mode. This model was used for the prediction of the formation cross sections of neutron-rich nuclides in fission.

Rubchenya, V. A.

2001-10-01

159

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

NASA Astrophysics Data System (ADS)

Short-lived fission gas release from fuel pellets during irradiation was investigated based on the experimental results of the gas-flow rigs irradiated in the Halden Heavy Water Reactor (HBWR). The release-to-birth ( R/ B) rates of short-lived fission gas were measured by means of gas-flow measurement during the irradiation experiments. Surface-to-volume ( S/ V) ratios of fuel pellets and diffusion coefficients of short-lived fission gas release were evaluated from the obtained ( R/ B) values. The increase of ( S/ V) ratio agreed well with the point where the fuel temperature exceeded the threshold of 1% fission gas release. This indicates that the interlinkage of fission gas bubbles occurred there. The evaluated diffusion coefficients scattered in the range between 10 -23 and 10 -17 m 2/s, and the effects of fuel type (UO 2 or MOX) were not clearly observed. In addition, it is likely that the restructuring effect of fuel pellet on the diffusion coefficients of short-lived fission gas at least in the fuel pellet matrix is negligible in high burnup region where the rim structure forms in the fuel pellet.

Amaya, Masaki; Grismanovs, Viktors; Tverberg, Terje

2010-07-01

160

Mass spectrometry studies of fission product behavior: 1, Fission products released from irradiated LWR (light-water reactor) fuel  

SciTech Connect

The chemical form and rate of release of volatile fission products (i.e., Xe, Kr, Cs, Te, I...) effused from an irradiated LWR fuel pin sample were studied using quadrupole mass spectrometry. Experiments, up to a temperature of 2120 K, 2060 K have identified krypton, xenon, cesium, and tellurium as the species released from the fuel. In addition, there was a weak signal for atomic iodine at 1325 K. The source of the atomic iodine, e.g. dissociation of cesium iodine or dissociation of molecular iodine, has yet to be resolved. The observed rate of release of xenon was several orders of magnitude lower than previously reported. However, the xenon release rate increased significantly after the fuel was oxidized. In complementary experiments on nonradioactive material, the release of tellurium was hindered by reaction with Zircaloy cladding. Above 1300/sup 0/C, gaseous SnTe was observed; its formation is attributed to reaction of the tin (in the cladding) with ZrTe/sub 2/. 4 refs., 5 figs.

Johnson, I.; Johnson, C.E.

1987-01-01

161

Prompt ?-ray production in neutron-induced fission of 239Pu  

NASA Astrophysics Data System (ADS)

Background: The prompt gamma-ray spectrum from fission is important for understanding the physics of nuclear fission, and also in applications involving fission. Relatively few measurements of the prompt gamma spectrum from 239Pu(n,f) have been published.Purpose: This experiment measured the multiplicity, individual gamma energy spectrum, and total gamma energy spectrum of prompt fission gamma rays from 239Pu(n,f) in the neutron energy range from thermal to 30 keV, to test models of fission and to provide information for applications.Method: Gamma rays from neutron-induced fission of 239Pu were measured using the DANCE gamma-ray calorimeter. Fission events were tagged by detecting fission products in a parallel-plate avalanche counter in the center of DANCE. The measurements were corrected for detector response using a geant4 model of DANCE. A detailed analysis for the gamma rays from the 1+ resonance complex at 10.93 eV is presented.Results: A six-parameter analytical parametrization of the fission gamma-ray spectrum was obtained. A Monte Carlo Hauser-Feshbach calculation provided good general agreement with the data, but some differences remain to be resolved.Conclusions: An analytic parametrization can be made of the gamma-ray multiplicity, energy distribution, and total-energy distribution for the prompt gamma rays following neutron-induced fission of 239Pu. This parametrization may be useful for applications. Modern Monte Carlo Hauser-Feshbach calculations can do a good job of calculating the fission gamma-ray emission spectrum, although some details remain to be understood.

Ullmann, J. L.; Bond, E. M.; Bredeweg, T. A.; Couture, A.; Haight, R. C.; Jandel, M.; Kawano, T.; Lee, H. Y.; O'Donnell, J. M.; Hayes, A. C.; Stetcu, I.; Taddeucci, T. N.; Talou, P.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Chyzh, A.; Gostic, J.; Henderson, R.; Kwan, E.; Wu, C. Y.

2013-04-01

162

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

NASA Astrophysics Data System (ADS)

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

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

2008-11-01

163

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

SciTech Connect

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.

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

2006-06-01

164

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

SciTech Connect

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.

Deforest, D.L.

1991-12-01

165

Natural gas production from Arctic gas hydrates  

Microsoft Academic Search

The natural gas hydrates of the Messoyakha field in the West Siberian basin of Russia and those of the Prudhoe Bay-Kuparuk River area on the North Slope of Alaska occur within a similar series of interbedded Cretaceous and Tertiary sandstone and siltstone reservoirs. Geochemical analyses of gaseous well-cuttings and production gases suggest that these two hydrate accumulations contain a mixture

1993-01-01

166

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

DOEpatents

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.

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

2007-10-02

167

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

SciTech Connect

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.

Lewis, J. M., E-mail: lewisj@ufl.edu; Kelley, R. P.; Jordan, K. A. [Nuclear Engineering Program, University of Florida, Gainesville, Florida 32611 (United States); Murer, D. [Arktis Radiation Detectors Ltd., 8045 Zurich (Switzerland)

2014-07-07

168

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

SciTech Connect

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.

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

1999-02-17

169

DFT-based prediction of fission product sorption on carbon structures under O2 ingress conditions  

NASA Astrophysics Data System (ADS)

An isotherm based model for the prediction of Cs sorption on the carbon components of a High Temperature Reactor (HTR) under O2 ingress conditions is presented. Isotherms are derived from a thermodynamic model based on binding energies calculated using Density Functional Theory (DFT). The DFT derived isotherms are compared with isotherms obtained from experimental calculations and sources of discrepancies are discussed. A DFT only model and a second model combining DFT and experimental calculations are used to predict fission product inventories in a HTR vessel during O2 ingress conditions. Results suggest that the carbon type (i.e. graphitic vs. amorphous) plays a central role on fission product sorption and release. During normal reactor conditions (T around 1400 K, low P) graphitic carbon will absorb a small percentage of a monolayer of Cs, while amorphous carbon will be approximately saturated at an entire monolayer of Cs. Results also indicate that, for the case of O2 ingress to the reactor's vessel, the Cs will form Cs2O. In the case of graphitic carbon, the Cs2O will bind more weakly than Cs, leading to Cs release in the form of Cs2O during O ingress. However, the weak binding of Cs to graphite means that only small release is expected. In the case of amorphous carbon, Cs2O binds almost as strongly Cs, and so no significant change in Cs absorbed to the amorphous carbon is predicted, although the form of the absorbed Cs is predicted to be Cs2O. For the case of low release conditions, consistent with modern TRISO fuels, the core will adsorb the entire Cs inventory at normal operating temperatures. However, for high Cs release conditions, consistent with older TRISO fuels, the surface sites on the core will be saturated and most of the Cs will remain in gas form or plate out on other surfaces.

Londono-Hurtado, Alejandro; Szlufarska, Izabela; Morgan, Dane

2013-06-01

170

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

171

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

SciTech Connect

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.

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

2013-09-01

172

Nuclide yields of light fission products from thermal-neutron induced fission of 233U at different kinetic energies  

NASA Astrophysics Data System (ADS)

The yields of light fission products from thermal-neutron induced fission of 233U are measured as a function of their mass A, their nuclear charge Z, their kinetic energy E and their ionic charge state q at the recoil spectrometer Lohengrin of the Institut Laue-Langevin in Grenoble. The mass yields are determined by intercepting the fragments with an ionization chamber of high energy resolution positioned at the focal plane of the spectrometer. The nuclear charges and their yields are determined with the same ionization chamber by measuring the residual energy of fission products, selected monoenergetically by Lohengrin, behind a passive absorber made of parylene-C. The nuclear charge resolution enabled by this detector device is considerably improved to Z/d Z = 58. The nuclear charge and mass distributions summed over all ionic charge states are listed within the mass range 79 ? A ? 106 at 6 energies: E = 85.34, 90.41, 95.46, 100.50, 105.55 and 110.55 MeV. The energy-integrated nuclear charge and mass yields are also given. The isotonic and isotopic yields are shown. An odd-even effect in the yields is found for the protons as well as for the neutrons at all kinetic energies. The yield weighted total odd-even effect for the protons is found to be (22.1 2.1)%, for the neutrons (5.4 1.7)%. An odd-even effect for the protons in the mean kinetic energy is also observed. The displacement of the mean isobaric nuclear charges from the unchanged charge-density values and the variances of the isobaric nuclear-charge distributions reveal fine structures in their mass dependences.

Quade, U.; Rudolph, K.; Skorka, S.; Armbruster, P.; Clerc, H.-G.; Lang, W.; Mutterer, M.; Schmitt, C.; Theobald, J. P.; Gnnenwein, F.; Pannicke, J.; Schrader, H.; Siegert, G.; Engelhardt, D.

1988-10-01

173

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

SciTech Connect

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

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

2005-10-01

174

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

SciTech Connect

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.

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

2006-05-09

175

Fission Product Immobilisation in Secondary Phases Formed During Magnox Waste Glass Dissolution at 60 C: Experimental Results and Modelling.  

E-print Network

Fission Product Immobilisation in Secondary Phases Formed During Magnox Waste Glass Dissolution]. However, far less research effort has been directed towards identifying the secondary products of glass a complex mixture of fission product oxides representative of HLW glasses resulting from Magnox fuel

Sheffield, University of

176

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

SciTech Connect

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.

Morris, Robert Noel [ORNL

2008-03-01

177

Structure of the mass distributions of the products of thermal-neutron fission of ²³⁵U and ²³⁹Pu for a fixed energy of the fission products  

Microsoft Academic Search

Relative mass yields of the fission products of thermal-neutron fission of ²³⁵U and ²³⁹Pu have been determined for six values of the kinetic energy E\\/sub k\\/. A fine structure is observed at A = 137 and an anomalously low yield at A = 139. Mass distributions of fission products of the heavy group are presented for a fixed value of

A. D. Belyaev; Z. S. Bikbova; V. I. Kogan; A. I. Muminov; V. P. Pikul; A. M. Usmandiyarov

1984-01-01

178

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

SciTech Connect

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

Nagy, B.; Rigali, M.J. (Univ. of Arizona, Tucson (United States))

1993-01-01

179

Statistical estimation of physical quantities in thermal- and fast-neutron-induced fission. [Fission product mass yields, fragment kinetic energies, numbers of prompt neutrons  

Microsoft Academic Search

Making use of a model based on the statistical theory in which the scission-point distance is treated as an adjustable parameter, calculations were performed to obtain the mass yields of fission products, the kinetic energies of fission fragments and the numbers of prompt neutrons from neutron-induced fission of ²³²Th, ²³¹Pa, ²³³U, ²³⁵U, ²³⁸U, ²³⁷Np, ²³⁹Pu and ²⁴¹Pu for incident-neutron energies

T. Yamamoto; K. Sugiyama

1975-01-01

180

Trapping and diffusion of fission products in ThO2 and CeO2  

SciTech Connect

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.

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

2011-01-01

181

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

SciTech Connect

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.

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

1991-11-01

182

Spent Nuclear Fuel project estimate of volatile fission products release from multi-canister overpacks  

SciTech Connect

Spent N-Reactor fuel will be moved from wet pool storage to dry storage at Hanford Washington. This fuel will be sequentially loaded into a Multiple Container Overpack (MCO), moved to the cold vacuum drying station, drained, cold vacuum dried, shipped to the Canister Storage Building (CSB), staged for up to 2 years,hot vacuum dried at 300 degrees C, hot conditioned at 150 degrees C, and finally, sealed and stored for up to 75 years in the CSB.During each proposed process step, the volatile radioactive fission products released to the atmosphere were estimated.Tritium is the only volatile fission product released insignificant amounts during each process step. For an accident scenario involving interior MCO temperature of 600 degrees C for up to 8 hours, it was estimated that many volatile fission products are released.

Cooper, T.D.

1996-08-01

183

Results of fission products ? decay properties measurement performed with a total absorption spectrometer  

NASA Astrophysics Data System (ADS)

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

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

2014-03-01

184

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media  

NASA Astrophysics Data System (ADS)

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

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

2005-12-01

185

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

SciTech Connect

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.

Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

2014-09-01

186

Integral data testing of ENDF/B fission-product data and comparisons of ENDF/B with other fission product data files  

SciTech Connect

Three experiments (one from Oak Ridge and two from Los Alamos), in which samples of /sup 235/U and /sup 238/Pu were irradiated with thermal neutrons and either the total, gamma-ray, or gamma- and beta-ray fission product decay-energies were measured as functions of cooling time, were selected for comparisons with calculations made using four different fission product data files. The data files used were (1) the ENDF/B-IV fission product file, (2) the ENDF/B-V fission product file, (3) a file derived by substituting decay energies from JNDC into the ENDF/B-V file, and (4) a file derived by substituting decay-energies and spectra from the UK data file into the ENDF/B-V file. Direct summation calculations and spectral comparisons of the experiments were made using these data files as input, and both types of calculational analyses yielded the same results; namely, all data files are deficient, but the JNDC-ENDF/B-V results for the gamma- and beta-ray total decay-energy agree best with experiments. In addition, spectral comparisons with experiment generally indicate that calculated gamma-ray decay-energies are relatively high for early cooling times and small gamma-ray energies; they are low for early cooling times and large gamma-ray energies. The opposite is somewhat the case for the beta-ray decay energies; that is, the calculations are generally low for small beta-ray energies and high for large energies.

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

1981-11-01

187

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

SciTech Connect

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

Blaise Collin

2001-10-01

188

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

SciTech Connect

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

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

1984-01-01

189

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

NASA Astrophysics Data System (ADS)

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.

Yun, Di

190

Nuclear Power from Fission Reactors. An Introduction.  

ERIC Educational Resources 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

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

191

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

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

192

Shale gas production: potential versus actual greenhouse gas emissions  

E-print Network

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 ...

OSullivan, Francis Martin

193

Four-Fold Data Analysis of 252Cf Fission Products  

NASA Astrophysics Data System (ADS)

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

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

194

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

SciTech Connect

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.

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

2013-05-15

195

Revision of fission product nuclear data for JENDL-3.2  

SciTech Connect

Reevaluation of the nuclear data for about 60 fission product nuclides has been made for JENDL-3.2, taking account of new experimental data and results of integral tests for JENDL-3.l. Integral tests of the revised data were made with the STEK experiments.

Kawai, Masayoshi [Toshiba Corp., Kawasaki (Japan); Nakagawa, Tsuneo; Chiba, Satoshi; Nakajima, Yutaka; Sugi, Teruo; Kikuchi, Yasuyuki [Japan Atomic Energy Res. Inst., Ibaraki-ken (Japan); Zukeran, Atsushi [Hitachi Ltd., Hitachi-shi (Japan); Watanabe, Takashi [Kawasaki Heavy Industries, Ltd., Tokyo (Japan); Matsunobu, Hiroyuki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)

1994-12-31

196

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

Microsoft Academic Search

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

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

2011-01-01

197

November, 1967 Riso Report No. 170 Investigations on the Plant Uptake of Fission Products from Contaminated  

E-print Network

Contaminated Soils. I. Influence of Plant Species and Soil Types on the Uptake of Radioactive StrontiumNovember, 1967 Riso Report No. 170 Investigations on the Plant Uptake of Fission Products from plant species are presented, and comparisons are made between the uptake figures registered on some

198

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...

199

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

SciTech Connect

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

Gibbs, A.; Thomason, R.S.

2000-09-05

200

Isomeric yield ratio of fission product148Pr in235U( n th, f)  

NASA Astrophysics Data System (ADS)

The independent isomeric yield ratio of148Pr in thermal neutron induced fission of235U has been determined experimentally. The fission product148Pr isomers, extracted directly by on-line mass separation technique, have high-spin ( J=4) to low-spin ( J=1) isomer ratio of 0.140.04 using growth and decay analysis. Statistical model calculation of isomeric yeild ratio using constant initial r.m.s. angular momentum J rms can not reproduce either present results or other recent measurements of isomer ratios. The J rms derived from isomer ratio data in all thermal fissioning systems indicate a wide spread ranging from 2 ? to 13 ?. No clear correlation between J rms and isomeric spins or number of neutrons of isomers is found, thus, more model refinements and experimental works should be done in order to evaluate independent isomeric yields correctly.

Chung, C.; Yuan, Liq-Ji; Walters, W. B.

1984-10-01

201

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

SciTech Connect

A fuel rod has been instrumented with a new design of an acoustic resonator used to measure in a non destructive way the internal rod plenum gas mixture composition. This ultrasonic sensor has demonstrated its ability to operate in pile during REMORA 3 irradiation experiment carried out in the OSIRIS Material Testing Reactor (CEA Saclay, France). Due to very severe experimental conditions such as temperature rising up to 150 deg.C and especially, high thermal fluence level up to 3.5 10{sup 19} n.cm{sup 2}, the initial sensor gas speed of sound efficiency measurement was strongly reduced due to the irradiation effects on the piezo-ceramic properties. Nevertheless, by adding a differential signal processing method to the initial data analysis procedure validated before irradiation, the gas resonance peaks were successfully extracted from the output signal. From these data, the molar fractions variations of helium and fission gas were measured from an adapted Virial state equation. Thus, with this sensor, the kinetics of gas release inside fuel rods could be deduced from the in-pile measurements and specific calculations. These data will also give information about nuclear reaction effect on piezo-ceramics sensor under high neutron and gamma flux. (authors)

Rosenkrantz, E.; Ferrandis, J. Y.; Augereau, F. [CNRS - Univ. Montpellier 2, Southern Electronic Inst., UMR 5214, F-34095 Montpellier (France); Lambert, T. [CEA DEN - Nuclear Energy Direction - Fuel Studies Dept. - Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Fourmentel, D. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, F-13108 Saint Paul-Lez-Durance (France); Tiratay, X. [CEA DEN, Nuclear Energy Div., Nuclear Reactors and Facilities Dept., F-91191 Gif Sur Yvette (France)

2011-07-01

202

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

SciTech Connect

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

Blaise Collin

2014-09-01

203

Application of adjusted data in calculating fission-product decay energies and spectra  

Microsoft Academic Search

The code ADENA, which approximately calculates fussion-product beta and gamma decay energies and spectra in 19 or fewer energy groups from a mixture of U235 and Pu239 fuels, is described. The calculation uses aggregate, adjusted data derived from a combination of several experiments and summation results based on the ENDF\\/B-V fission product file. The method used to obtain these adjusted

D. C. George; R. J. Labauve; T. R. England

1982-01-01

204

Natural gas production verification tests  

SciTech Connect

This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) in compliance with the requirements of the National Environmental Policy Act of 1969. The Department of Energy (DOE) proposes to fund, through a contract with Petroleum Consulting Services, Inc. of Canton, Ohio, the testing of the effectiveness of a non-water based hydraulic fracturing treatment to increase gas recovery from low-pressure, tight, fractured Devonian Shale formations. Although Devonian Shales are found in the Appalachian, Michigan, and Illinois Basins, testing will be done only in the dominant, historical five state area of established production. The objective of this proposed project is to assess the benefits of liquid carbon dioxide (CO{sub 2})/sand stimulations in the Devonian Shale. In addition, this project would evaluate the potential nondamaging (to the formation) properties of this unique fracturing treatment relative to the clogging or chocking of pores and fractures that act as gas flow paths to the wellbore in the target gas-producing zones of the formation. This liquid CO{sub 2}/sand fracturing process is water-free and is expected to facilitate gas well cleanup, reduce the time required for post-stimulation cleanup, and result in improved production levels in a much shorter time than is currently experienced.

Not Available

1992-02-01

205

Mass yield distributions of fission products from photo-fission of 238U induced by 11.5-17.3 MeV bremsstrahlung  

NASA Astrophysics Data System (ADS)

The yields of various fission products in the 11.5, 13.4, 15.0 and 17.3 MeV bremsstrahlung-induced fission of 238U have been determined by recoil catcher and an off-line ?-ray spectrometric technique using the electron linac, SAPHIR at CEA, Saclay, France. The mass yield distributions were obtained from the fission product yields using charge-distribution corrections. The peak-to-valley ( P/ V ratio, average light mass (< A L>) and heavy mass (< A H>) and average number of neutrons (< v>) in the bremsstrahlung-induced fission of 238U at different excitation energies were obtained from the mass yield data. From the present and literature data in the 238U ( ?, f ) and 238U ( n, f ) reactions at various energies, the following observations were obtained: i) The mass yield distributions in the 238U ( ?, f ) reaction at various energies of the present work are double-humped, similar to those of the 238U ( n, f ) reaction of comparable excitation energy. ii) The yields of fission products for A = 133-134, A = 138-140, and A = 143-144 and their complementary products in the 238U ( ?, f) reaction are higher than other fission products due to the nuclear structure effect. iii) The yields of fission products for A = 133-134 and their complementary products are slightly higher in the 238U ( ?, f ) than in the 238U ( n, f ) , whereas for A = 138-140 and 143-144 and their complementary products are comparable. iv) With excitation energy, the increase of yields of symmetric products and the decrease of the peak-to-valley ( P/ V ratio in the 238U ( ?, f) reaction is similar to the 238U ( n, f) reaction. v) The increase of < v> with excitation energy is also similar between the 238U ( ?, f ) and 238U ( n, f) reactions. However, it is surprising to see that the < A L> and < A H> values with excitation energy behave entirely differently from the 238U ( ?, f ) and 238U ( n, f ) reactions.

Naik, H.; Carrel, Frdrick; Kim, G. N.; Laine, Frdric; Sari, Adrien; Normand, S.; Goswami, A.

2013-07-01

206

Application of adjusted data in calculating fission-product decay energies and spectra  

NASA Astrophysics Data System (ADS)

The code ADENA, which approximately calculates fussion-product beta and gamma decay energies and spectra in 19 or fewer energy groups from a mixture of U235 and Pu239 fuels, is described. The calculation uses aggregate, adjusted data derived from a combination of several experiments and summation results based on the ENDF/B-V fission product file. The method used to obtain these adjusted data and the method used by ADENA to calculate fission-product decay energy with an absorption correction are described, and an estimate of the uncertainty of the ADENA results is given. Comparisons of this approximate method are made to experimental measurements, to the ANSI/ANS 5.1-1979 standard, and to other calculational methods. A listing of the complete computer code (ADENA) is contained in an appendix. Included in the listing are data statements containing the adjusted data in the form of parameters to be used in simple analytic functions.

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

1982-06-01

207

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

SciTech Connect

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

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

1980-01-01

208

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

SciTech Connect

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

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

1980-01-01

209

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

SciTech Connect

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

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

2014-04-07

210

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

E-print Network

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 ...

Greenwood, R C; Novick, V J

1981-01-01

211

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

SciTech Connect

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

Robin D. Rogers

2004-12-09

212

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

SciTech Connect

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

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

1991-01-01

213

SCRAM-discharge volume break studies. Part 2. Fission-product transport analyses. [PWR; BWR  

SciTech Connect

This portion of the ORNL-SASA presentation deals with the analysis of the rate of movement of fission products from the overheated core through a series of reactor control volumes, the final one being the exterior of the reactor building. At this time, the analysis of a complete station blackout sequence at Browns Ferry has been completed. The fission product transport portion of the study was presented in preliminary form at the 1981 Water Reactor Safety Meeting. Currently, the analysis of the small-break LOCA outside of the containment is in process. The initial study traced noble gases and iodine through the reactor systems during the event sequence; the current work includes an analysis of cesium transport in addition to noble gases and iodine.

Wichner, R.P.; Weber, C.F.; Lorenz, R.A.; Nehls, J.W.; Wright, A.L.

1982-01-01

214

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

SciTech Connect

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

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

1984-01-01

215

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

SciTech Connect

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

Lawrence, G.P.

1991-01-01

216

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

Microsoft Academic Search

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

Masaki Ozawa; Yoshihiko Shinoda; Yuichi Sano

2002-01-01

217

Correlations for fission product release from N Reactor fuel under high-temperature accident conditions  

SciTech Connect

Empirical correlations were derived for fission product release from metallic uranium alloy 601 N Reactor fuel during postulated accident conditions in which the fuel nears, reaches, or exceeds the melting temperature. The correlations were based on a sparse data base from fuel melted in an inert or steam atmosphere. The empirical correlations are presented for use in subsequent deterministic analyses of N Reactor behavior during hypothetical severe accidents beyond the design basis. 20 refs., 4 figs., 4 tabs.

Birney, K.R.; Bechtold, D.B.; McCall, T.B.

1988-03-01

218

Fission-Product Separation Based on Room-Temperature Ionic Liquids  

SciTech Connect

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.

Luo, Huimin; Hussey, Charles L.

2005-09-30

219

Fission-Product Separation Based on Room-Temperature Ionic Liquids  

SciTech Connect

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.

Luo, Huimin

2006-11-15

220

Effects of time and other variables on fission product release rates  

Microsoft Academic Search

The releases of krypton and cesium from highly irradiated LWR fuel have been examined in detail. The main interest has been the effect of time on the rate of release and the effects of heatup and cooldown cycles. The minute-by-minute release rates for fission product ⁸⁵Kr from commercial fuel irradiated in the H.B. Robinson PWR are shown. The release rate,

R. A. Lorenz; M. F. Osborne; Jack Lee Collins

1986-01-01

221

Neutronics analysis of water-cooled energy production blanket for a fusionfission hybrid reactor  

Microsoft Academic Search

Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusionfission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1GWe with self-sustaining tritium, i.e. the M

Jieqiong Jiang; Minghuang Wang; Zhong Chen; Yuefeng Qiu; Jinchao Liu; Yunqing Bai; Hongli Chen; Yanglin Hu

2010-01-01

222

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

SciTech Connect

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.

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

2014-07-07

223

Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization  

SciTech Connect

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.

Joshua R. Versey

2013-08-01

224

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

PubMed

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

Abrecht, David G; Schwantes, Jon M

2015-03-01

225

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

SciTech Connect

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.

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

1988-01-01

226

Application of adjusted data in calculating fission-product decay energies and spectra. [ADENA code  

SciTech Connect

The code ADENA, which approximately calculates fission-product beta and gamma decay energies and spectra in 19 or fewer energy groups from a mixture of /sup 235/U and /sup 239/Pu fuels, is described. The calculation uses aggregate, adjusted data derived from a combination of several experiments and summation results based on the ENDF/B-V fission-product file. The method used to obtain these adjusted data and the method used by ADENA to calculate fission-product decay energy with an absorption correction are described, and an estimate of the uncertainty of the ADENA results is given. Comparisons of this approximate method are made to experimental measurements, to the ANSI/ANS 5.1-1979 standard, and to other calculational methods. A listing of the complete computer code (ADENA) is contained in an appendix. Included in the listing are data statements containing the adjusted data in the form of parameters to be used in simple analytic functions. These fitted parameters can be abstracted for other uses such as in spatial neutron depletion or thermal hydraulics codes.

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

1982-06-01

227

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

SciTech Connect

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

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

1988-01-01

228

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

PubMed

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

Mushtaq, A

2011-04-01

229

HYPER-FUSE - A novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation  

NASA Astrophysics Data System (ADS)

A conservative, simplified analytical model is adapted to carry out the parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR. The principal parameters of interest are mentioned. Other fission products of possible interest for transmutation are analyzed. Possible reactor design for hyper-fuse are examined and the rail gun accelerator is found promising for propelling the pellets to the required velocities.

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

230

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

Microsoft Academic Search

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

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

1979-01-01

231

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

NASA Astrophysics Data System (ADS)

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

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

2014-05-01

232

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

SciTech Connect

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

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

2014-05-01

233

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

NASA Astrophysics Data System (ADS)

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

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

2014-03-01

234

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

SciTech Connect

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.

James Stubbins

2012-12-19

235

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

NASA Astrophysics Data System (ADS)

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.

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

2013-05-01

236

Characterization of the LISOL laser ion source using spontaneous fission of 252Cf  

NASA Astrophysics Data System (ADS)

A spontaneous fission Californium-252 source was placed inside a gas cell in order to characterize the LISOL laser ion source. The fission products from 252Cf are thermalized and neutralized in the plasma created by energetic particles. Two-step selective laser ionization is applied to produce purified beams of radioactive isotopes. The survival of fission products in a single charge state has been studied in argon as a buffer gas for different elements.

Kudryavtsev, Yu.; Cocolios, T. E.; Gentens, J.; Ivanov, O.; Huyse, M.; Pauwels, D.; Sawicka, M.; Sonoda, T.; Van den Bergh, P.; Van Duppen, P.

2008-10-01

237

Production Trends of Shale Gas Wells  

E-print Network

To obtain better well performance and improved production from shale gas reservoirs, it is important to understand the behavior of shale gas wells and to identify different flow regions in them over a period of time. It is also important...

Khan, Waqar A.

2010-01-14

238

Ionization of noble gas atoms by alpha particles and fission fragments from the decay of 252Cf1  

NASA Astrophysics Data System (ADS)

Charge state distributions of He, Ne, Ar, Kr, and Xe ions produced in single collisions with alpha particles and fission fragments from the decay of 252Cf have been measured using time of flight spectrometry. The measurements reveal that the maximum number of electrons removed in a fission fragment collision ranges from eight in the case of Ne to 20 in the case of Xe. Recoil-ion production cross sections have been determined for the resolvable ionic charge states and compared with the predictions of a model based upon the independent electron approximation.

Hill, B. M.; Watson, R. L.; Wohrer, K.; Bandong, B. B.; Sampoll, G.; Horvat, V.

1993-07-01

239

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

DOEpatents

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.

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

2009-01-06

240

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

DOEpatents

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.

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

2009-05-05

241

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

DOEpatents

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.

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

2009-01-27

242

Mass-yield distribution of the fission products in fallout from the 14 may 1965 nuclear explosion.  

PubMed

Twenty single particles separated from a 20-liter sample of rain collected in Osaka, Japan, shortly after the 14 May 1965 test explosion of the Chinese nuclear device, were analyzed radiochemically. The abundance pattern of the fission products in these particles resembled the shape of the mass-yield curve for the thermal neutron-induced fission of uranium-235, except for the facts that cesium-hl37 and strontium-90 were markedly depleted and the yields near the symmetric fission region appeared to be somewhat enhanced. PMID:17757237

Rao, M N; Yoshikawa, K; Sabu, D D; Clark, R; Kuroda, P K

1966-08-01

243

Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges  

SciTech Connect

The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations of the noble metals in HLW sludges are measured by using mass spectroscopy to determine concentrations of the isotopes that comprise each noble metal. For example, the noble metal Ru is comprised of isotopes with masses 101, 102, and 104. The element Rh has a single isotope with mass 103. The element Pd is comprised of five isotopes. These are at masses 105-108 and mass 110. As does Rh, Ag has only one isotope. This is at mass 109. However, results in this report show that the Ag concentration in the two samples was due to natural Ag being in the samples. Natural Ag has masses at 107 and 109. The Ag-107 interferes with the measurement of Pd-107. This Ag was used in one of the processes at SRS. The results also show that natural Cd is in the two samples. Cadmium has isotopes at masses 106, 108 and 110, thus it interferes with the analysis of the Pd isotopes at these masses. Cadmium was also used in one of the processes at SRS. However, the concentrations of the Pd isotopes at masses 106, 107, 108 and 110 could be calculated using the fission yields for the Pd isotopes, and the measured concentration of Pd at mass 105 where there is no Ag or Cd interference. Based on the measurements of the concentrations of the isotopes of each noble metal, the total concentration of that noble metal can be determined by summing the concentrations of the individual isotopes. The results in this report show that the relative concentrations of the isotopes of Ru and Rh are in proportion to their yields from the fission of U-235 in the reactors. These results were expected since these elements are very insoluble in caustic and thus are primarily in the sludge tanks rather then the salt tanks of the SRS Tank Farm. The relative concentration of Pd is somewhat lower than that based on the relative fission yields of its five isotopes. This indicates that some of the Pd is in the salt tanks rather than the sludge tanks of the Tank Farm. The concentrations of the noble metals were predicted using the High Level Waste Characterization System (WCS) at SRS. This system keeps record of the inventory of the major compounds and select radionuclides that are in each of the SRS HLW tanks. Using this system, the Closure Business Unit (CBU) can predict the major composition of a sludge ba

Bibler, N

2005-04-05

244

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

NASA Astrophysics Data System (ADS)

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

Tipnis, Sameer Vijay

245

Gas production in distant comets  

NASA Astrophysics Data System (ADS)

Molecular spectroscopy at radio wavelengths is a tool well suited for studying the composition and outgassing kinematics of cometary comae. This is particularly true for distant comets, i.e. comets at heliocentric distances greater than a few AU, where the excitation of molecules is inefficient other than for rotational energy levels. At these distances, water sublimation is inefficient, and cometary activity is dominated by outgassing of carbon monoxide. An observing campaign is presented, where the millimeter- wave emission from CO in comet 29P/Schwassmann-Wachmann 1 has been studied in detail using the Swedish-ESO Submillimetre Telescope (SEST). Coma models have been used to analyse the spectra. The production of CO is found to have two separate sources, one releasing CO gas on the nuclear dayside, and one extended source, where CO is produced from coma material, proposed to be icy dust grains. Radio observations of many molecules in comet C/1995 O1 (Hale-Bopp) have been carried out in a long-term international effort using several radio telescopes. An overview of the results is presented, describing the evolution of the gas production as the comet passed through the inner Solar system. Spectra recorded using the SEST, primarily of CO, for heliocentric distances from 3 to 11 AU are analysed in detail, also using coma models. The concept of icy grains constituting the extended source discovered in comet 29P/Schwassmann-Wachmann 1 is examined by theoretical modelling of micrometre-sized ice/dust particles at 6 AU from the Sun. It is shown that that such grains can release their content of volatiles on timescales similar to that found for the extended source.

Gunnarsson, Marcus

246

Isomer production ratios and the angular momentum distribution of fission fragments  

NASA Astrophysics Data System (ADS)

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.

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

2013-10-01

247

Delayed beta- and gamma-ray production due to thermal-neutron fission of ²³⁵U, spectral distributions for times after fission between 2 and 14,000 sec: tabular and graphical data  

Microsoft Academic Search

Fission-product decay energy-release rates were measured for thermal-neutron fission of ²³⁵U. Samples of mass 1 to 10 ..mu..g were irradiated for 1 to 100 s by using the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were counted for times-after-fission between 2 and 14,000 s. The data were obtained for beta and gamma

J. K. Dickens; T. A. Love; J. W. McConnell; J. F. Emery; K. J. Northcutt; R. W. Peelle; H. Weaver

1978-01-01

248

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

NASA Astrophysics Data System (ADS)

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

Rest, J.

2004-03-01

249

Simulation of the effects of grain boundary fission gas during thermal transients  

SciTech Connect

This report presents the results of an initial set of out-of-cell transient heating experiments performed on unirradiated UO/sub 2/ pellets fabricated to simulate the effect of grain boundary fission gas on fuel swelling and cladding failure. The fabrication involved trapping high-pressure argon on internal pores by sintering annular UO/sub 2/ pellets in a hot isostatic press (HIP). The pellet stack was subjected to two separate transients (DGF83-03A and -03B). Figures show photomicrographs of HIPped and non-HIPped UO/sub 2/, respectively, and the adjacent cladding after DGF83-03B. Fuel melting occurred at the center of both the HIPped and non-HIPped pellets; however, a dark ring is present near the center in the HIPped fuel but not in the non-HIPped fuel. This dark band is a high-porosity region due to increased grain boundary/edge swelling in that pellet. In contrast, grain boundary/edge swelling did not occur in the non-HIPped pellets. Thus, the presence of the high-pressure argon trapped on internal pores during sintering in the HIP altered the microstructural behavior. Results of these preliminary tests indicate that the microstructural behavior of HIPped fuel during thermal transients is different from the behavior of conventionally fabricated fuel.

Fenske, G.R.; Emerson, J.E.; Beiersdorf, B.A.

1984-11-01

250

COMPUTATION OF EARLY-TIME FISSION PRODUCT DOSE-RATE SPECTRA AND GAMMA-RAY AIR ATTENUATION  

Microsoft Academic Search

On the basis of photon spectra measurement for shorttime irradiations of ; U\\/sup 235, fission-product dose-rate spectra are computed for 1.7 to 1550 seconds ; after fission. Airattenuation curves that would result from point-isotropic ; sources having such spectral distributions are then coinputed. The fact that the ; attenuation curves are nearly straight lines when plotted against distance from ;

C. F. Ksanda; E. Lauments

1959-01-01

251

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

PubMed

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

Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

2012-04-01

252

HYPERFUSE: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation  

SciTech Connect

Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., /sup 137/Cs or /sup 90/Sr. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n, 2n), (n, ..cap alpha..), etc.) that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product.

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

1980-01-01

253

FISSION REACTORS KEYWORDS: high-temperature  

E-print Network

conversion system, and the progress in the electrolysis cell materials field can help the econom- icalFISSION REACTORS KEYWORDS: high-temperature steam electrolysis, advanced gas reactor, supercritical CO2 cycle HYDROGEN PRODUCTION USING HIGH-TEMPERATURE STEAM ELECTROLYSIS SUPPORTED BY ADVANCED GAS

Yildiz, Bilge

254

Fission-product behaviour in irradiated TRISO-coated particles: Results of the HFR-EU1bis experiment and their interpretation  

NASA Astrophysics Data System (ADS)

It is important to understand fission-product (FP) and kernel micro-structure evolution in TRISO-coated fuel particles. FP behaviour, while central to severe-accident evaluation, impacts: evolution of the kernel oxygen potential governing in turn carbon oxidation (amoeba effect and pressurization); particle pressurization through fission-gas release from the kernel; and coating mechanical resistance via reaction with some FPs (Pd, Cs, Sr). The HFR-Eu1bis experiment irradiated five HTR fuel pebbles containing TRISO-coated UO 2 particles and went beyond current HTR specifications (e.g., central temperature of 1523 K). This study presents ceramographic and EPMA examinations of irradiated urania kernels and coatings. Significant evolutions of the kernel (grain structure, porosity, metallic-inclusion size, intergranular bubbles) as a function of temperature are shown. Results concerning FP migration are presented, e.g., significant xenon, caesium and palladium release from the kernel, molybdenum and ruthenium mainly present in metallic precipitates. The observed FP and micro-structural evolutions are interpreted and explanations proposed. The effect of high flux rate and high temperature on fission-gas behaviour, grain-size evolution and kernel swelling is discussed. Furthermore, Cs, Mo and Zr behaviour is interpreted in connection with oxygen-potential. This paper shows that combining state-of-the-art post-irradiation examination and state-of-the-art modelling fundamentally improves understanding of HTR fuel behaviour.

Barrachin, M.; Dubourg, R.; de Groot, S.; Kissane, M. P.; Bakker, K.

2011-08-01

255

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)

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.

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

2014-07-01

256

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

NASA Astrophysics Data System (ADS)

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

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

2013-07-01

257

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

SciTech Connect

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.

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

1997-01-01

258

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

SciTech Connect

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

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

2005-09-01

259

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

SciTech Connect

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.

Murphy, B.D.; Dickens, J.K.; Walker, R.L.; Newton, T.D. [Oak Ridge National Lab., TN (United States)

1994-12-31

260

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

NASA Astrophysics Data System (ADS)

Platinoid fission products present in high-level nuclear wastes present particular challenges to their treatment by vitrification. The platinoid metals Ru, Rh, Pd, and their compounds are sparingly soluble in borosilicate glass melts. During glass melting under oxidizing conditions, the platinoids form small crystals of highly dense solid intermetallic phases and oxides. Under reducing conditions, the platinoids form only intermetallic phases. A fraction of these crystals settles to the bottom of the melting furnace, forming an immobile sludge. The fraction settling reported in the literature is highly variable. In the present work, the fraction settling was found to be >90% under reducing conditions but only 10 to 20% under oxidizing conditions. The thickness of the sludge layer depends on the volume fraction of platinoid crystals in the sludge, which is poorly known (typically ~0.06 under oxidizing conditions). Since the electrical conductivity of the sludge can be >10X that of the melt, in joule-heated melters the presence of such a layer can lead to diversion of the electric current, thereby compromising melter operability. The time to failure by this mechanism is clearly of practical importance. A variety of data are required in order to estimate the time to failure due to this mechanism and such data must be obtained under conditions representative of those in a full-size melting furnace. We have acquired such data using a melting furnace installed in our laboratory. This furnace is a one-third scale prototype of the system to be used for the vitrification of defense HLW at Hanford, WA. In the present work, simulated Hanford HLW material was combined with glass formers to produce a melter feed slurry that was then spiked with the platinoids. Over one thousand chemical and optical analyses were performed on hundreds of samples taken from the feed, various locations inside the furnace, the glass melt during pouring, the solid glass, and various locations along the prototypical off-gas treatment system. In the course of several weeks of testing, a total mass of 28,500 kg of glass was produced and sampled. The effect of operating conditions on the behavior of the platinoids was evaluated, including mixing the melt by bubbling with air vs. not bubbling, and the effects of reducing conditions (by adding sugar to the feed). Tests were conducted with Ru, Rh, Pd (0.17% total oxides) or Ru only (0.09 wt%) in the final glass product. The fractions of the platinoids discharged with the glass, deposited in the melter, and/or released to the off-gas were calculated from the analytical data. In addition, mathematical modeling of the distribution and movement of platinoid crystals within the melt was conducted for various furnace operating conditions. This modeling captured the flow, electrical, and thermal fields within the melt and included coupling of the local material properties to the local temperature. The experimental data on platinoid particle size and morphology were used to provide input for modeling their flow and sedimentation behavior with the objective of estimating accumulation rates and spatial distributions. The modeled deposition of the crystals was found to be uneven, with piles in the corners and thicker layers on slanted bottom surfaces. Consequently, contiguous electrical shorting paths could develop more quickly than what would be assumed based on uniform deposition. This paper will present the results from the experimental and modeling work and discuss their implications for melter lifetime estimation.

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

2006-05-01

261

Fission yields for thermal-neutron fission of plutonium-241  

Microsoft Academic Search

Cumulative fission yeilds for 17 fission products (16 mass chains) created by thermal-neutron fission of ²⁴¹Pu were determined from analysis of gross fission product gamma-ray spectra obtained using a large-volume Ge(Li) detector. Uncertainties assigned to nine of the measured yields are smaller than existing evaluated uncertainties. 25 references.

Dickens

1979-01-01

262

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

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.

Rest, J.

1995-08-01

263

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

Code of Federal Regulations, 2011 CFR

...false (Item 1204) Oil and gas production...prices and production costs. 229.1204 Section...1204 (Item 1204) Oil and gas production...prices and production costs. (a) For each...conversion to synthetic oil or gas, the product's...prices, and production costs should be...

2011-04-01

264

Breast cancer: evidence for a relation to fission products in the diet.  

PubMed

To establish the possible relation between breast cancer mortality and low doses of radiation due to fission products in the environment, the mortality rates in the nine census regions of the United States for the years 1984-1988 were correlated with the cumulative airborne releases from all the nuclear plants in each region for the period 1970-1987. A high correlation coefficient of 0.91 was obtained for a logarithmic dependence on the total releases, consistent with an indirect action via free-radical oxygen at very low dose rates, in contrast to a direct action on DNA at high dose rates, explaining the wide differences in risk per unit dose obtained in earlier studies. The recent temporal changes of breast cancer rates in the New York metropolitan area including nearby Connecticut, Westchester, and Long Island were examined in relation to the releases from nearby nuclear plants and found to be consistent with a dominant role of short-lived fission products in drinking water and fresh milk. The results support a major role for nuclear plant releases in industrial countries in the recent rises of breast and other forms of cancers not related to smoking, especially among older persons, and strongly support the need to replace nuclear reactors with more benign ways to generate electricity. PMID:8276535

Sternglass, E J; Gould, J M

1993-01-01

265

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

SciTech Connect

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.

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

1984-01-01

266

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

NASA Astrophysics Data System (ADS)

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.

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

2013-06-01

267

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

SciTech Connect

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.

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

2013-02-01

268

Inherent and Passive Safety Sodium-Cooled Fast Reactor Core Design with Minor Actinide and Fission Product Incineration  

Microsoft Academic Search

A self-consistent nuclear energy system (SCNES) can be a promising option as a future nuclear energy source. An SCNES should fulfill (a) efficient energy generation, (b) fuel production or breeding, (c) burning minor actinides with incinerating fission products, and (d) system safety. We focus on the system safety and present a simple evaluation model for the inherent and passive power

Hideaki Kuraishi; Tetsuo Sawada; Hisashi Ninokata; Hiroshi Endo

2001-01-01

269

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

NASA Technical Reports Server (NTRS)

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.

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

1975-01-01

270

Natural gas hydrates - issues for gas production and geomechanical stability  

E-print Network

in an offshore hydrate deposit. I modeled geomechanical failures associated with gas production using a horizontal well and a vertical well for two different types of sediments, sand and clay. The simulation results showed that the sediment and failures can... be a serious issue during the gas production from weaker sediments such as clays. v DEDICATION I dedicate this dissertation to my family; my mother and father, my brother, Arun and my sister, Aarti. It is only because of their love and support...

Grover, Tarun

2008-10-10

271

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

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.

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

2010-12-15

272

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

SciTech Connect

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.

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

2014-05-01

273

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

SciTech Connect

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.

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

2012-10-01

274

Ab Initio Molecular Dynamics of High-Temperature Unimolecular Dissociation of Gas-Phase RDX and Its Dissociation Products.  

PubMed

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

Schweigert, Igor V

2015-03-26

275

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

SciTech Connect

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.

Wright, R.Q.

1999-01-01

276

Seminar on Fission VI  

NASA Astrophysics Data System (ADS)

Topical reviews. Angular momentum in fission / F. Gnnenwein ... [et al.]. The processes of fusion-fission and quasi-fission of heavy and super-heavy nuclei / M. G. Itkis ... [et al.] -- Fission cross sections and fragment properties. Minor-actinides fission cross sections and fission fragment mass yields via the surrogate reaction technique / B. Jurado ... [et al.]. Proton-induced fission on actinide nuclei at medium energy / S. Isaev ... [et al.]. Fission cross sections of minor actinides and application in transmutation studies / A. Letourneau ... [et al.]. Systematics on even-odd effects in fission fragments yields: comparison between symmetric and asymmetric splits / F. Rejmund, M Caamano. Measurement of kinetic energy distributions, mass and isotopic yields in the heavy fission products region at Lohengrin / A. Bail ... [et al.] -- Ternary fission. On the Ternary [symbol] spectrum in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Energy degrader technique for light-charged particle spectroscopy at LOHENGRIN / A. Oberstedt, S. Oberstedt, D. Rochman. Ternary fission of Cf isotopes / S. Vermote ... [et al.]. Systematics of the triton and alpha particle emission in ternary fission / C. Wagemans, S. Vermote, O. Serot -- Neutron emission in fission. Scission neutron emission in fission / F.-J. Hambsch ... [et al.]. At and beyond the Scission point: what can we learn from Scission and prompt neutrons? / P. Talou. Fission prompt neutron and gamma multiplicity by statistical decay of fragments / S. Perez-Martin, S. Hilaire, E. Bauge -- Fission theory. Structure and fission properties of actinides with the Gogny force / H. Goutte ... [et al.]. Fission fragment properties from a microscopic approach / N. Dubray, H. Goutte, J.-P. Delaroche. Smoker and non-smoker neutron-induced fission rates / I. Korneev ... [et al.] -- Facilities and detectors. A novel 2v2E spectrometer in Manchester: new development in identification of fission fragments / I. Tsekhanovich ... [et al.]. Development of PSD and ToF + PSD techniques for fission experiments / M. Sillanp ... [et al.]. MYRRHA, a new fast spectrum facility / H. At Abderrahim, P. D'hondt, D. De Bruyn. The BR1 reactor: a versatile tool for fission experiments / J. Wagemans -- "Special" fission processes. Shape isomers - a key to fission barriers / S. Oberstedt ... [et al.]. Fission in spallation reactions / J. Cugnon, Th. Aoust, A. Boudard -- Conference photo -- List of participants.

Wagemans, Cyriel; Wagemans, Jan; D'Hondt, Pierre

2008-04-01

277

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

Code of Federal Regulations, 2014 CFR

...false (Item 1204) Oil and gas production...prices and production costs. 229.1204 Section...1204 (Item 1204) Oil and gas production...prices and production costs. (a) For each...computed using production costs disclosed pursuant to...Extractive ActivitiesOil and Gas....

2014-04-01

278

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

Code of Federal Regulations, 2013 CFR

...false (Item 1204) Oil and gas production...prices and production costs. 229.1204 Section...1204 (Item 1204) Oil and gas production...prices and production costs. (a) For each...computed using production costs disclosed pursuant to...Extractive ActivitiesOil and Gas....

2013-04-01

279

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

Code of Federal Regulations, 2012 CFR

...false (Item 1204) Oil and gas production...prices and production costs. 229.1204 Section...1204 (Item 1204) Oil and gas production...prices and production costs. (a) For each...computed using production costs disclosed pursuant to...Extractive ActivitiesOil and Gas....

2012-04-01

280

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

Code of Federal Regulations, 2010 CFR

... (Item 1204) Oil and gas production, production...product sold, of oil, gas, and other products...be made by geographical area and for each country...disclose, by geographical area: (1) The average...transfers) per unit of oil, gas and other products...

2010-04-01

281

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

SciTech Connect

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.

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

2014-01-01

282

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

PubMed

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

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

2012-11-16

283

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

NASA Astrophysics Data System (ADS)

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

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

2012-11-01

284

Measurement of Kinetic Energy Distributions, Mass and Isotopic Yields in the Heavy Fission Products Region at Lohengrin  

NASA Astrophysics Data System (ADS)

Mass yields and kinetic energy distribution functions for heavy mass fission products from thermal neutron induced fission of 235U and 239Pu have been measured at the spectrometer Lohengrin at the high flux reactor of the Institut Laue-Langevin in Grenoble. In addition to these measurements where an ionization chamber was used for the mass identification, we also performed gamma spectrometry to quantify the isotopic and isomeric yields. This setup using Ge-detectors has been commissioned with the system 241Pu(nth,f). In order to extend the data to less abundant fission products, a proportional counter for beta detection has been constructed, allowing to reduce the background by beta-gamma coincidences.

Bail, A.; Serot, O.; Litaize, O.; Faust, H. R.; Kster, U.; Materna, T.; Letourneau, A.; Dupont, E.

2008-04-01

285

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

SciTech Connect

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.

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

2007-02-28

286

Radioactive beams from {sup 252}CF fission using a gas catcher and an ECR charge breeder at ATLAS.  

SciTech Connect

An upgrade to the radioactive beam capability of the ATLAS facility has been proposed using {sup 252}CF fission fragments thermalized and collected into a low-energy particle beam using a helium gas catcher. In order to reaccelerate these beams an existing ATLAS ECR ion source will be reconfigured as a charge breeder source. A 1 Ci{sup 252}CF source is expected to provide sufficient yield to deliver beams of up to {approx}10{sup 6} far from stability ions per second on target. A facility description and the expected performance will be presented in this paper.

Savard, G.; Pardo, R. C.; Moore, E. F.; Hecht, A. A.; Baker, S.

2005-01-01

287

Radioactive beams from 252Cf fission using a gas catcher and an ECR charge breeder at ATLAS  

NASA Astrophysics Data System (ADS)

The Californium Rare Ion Breeder Upgrade (CARIBU) for the ATLAS facility is under construction. The facility will use 252Cf fission fragments thermalized and collected into a low-energy particle beam by a helium gas catcher. In order to reaccelerate these beams an existing ATLAS ECR ion source is being reconfigured as a charge breeder source. A 1Ci 252Cf source will provide sufficient yield to deliver beams of up to 10 6 far-from-stability ions per second on target. The facility design, expected performance and the project status will be presented in this paper.

Pardo, Richard C.; Savard, Guy; Baker, Sam; Davids, Cary; Moore, E. Frank; Vondrasek, Rick; Zinkann, Gary

2007-08-01

288

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

SciTech Connect

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.

Savard, Guy; Pardo, Richard C.; Moore, E. Frank; Hecht, Adam A.; Baker, Sam [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2005-03-15

289

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2014 CFR

...2014-04-01 2014-04-01 false Combustion product gas. 173.350 Section 173...Specific Usage Additives 173.350 Combustion product gas. The food additive combustion product gas may be safely used in...

2014-04-01

290

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2010 CFR

...2010-04-01 2009-04-01 true Combustion product gas. 173.350 Section 173...Specific Usage Additives 173.350 Combustion product gas. The food additive combustion product gas may be safely used in...

2010-04-01

291

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2013 CFR

...2013-04-01 2013-04-01 false Combustion product gas. 173.350 Section 173...Specific Usage Additives 173.350 Combustion product gas. The food additive combustion product gas may be safely used in...

2013-04-01

292

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2012 CFR

...2012-04-01 2012-04-01 false Combustion product gas. 173.350 Section 173...Specific Usage Additives 173.350 Combustion product gas. The food additive combustion product gas may be safely used in...

2012-04-01

293

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2011 CFR

...2011-04-01 2011-04-01 false Combustion product gas. 173.350 Section 173...Specific Usage Additives 173.350 Combustion product gas. The food additive combustion product gas may be safely used in...

2011-04-01

294

Monthly Natural Gas Gross Production Report  

EIA Publications

Monthly natural gas gross withdrawals estimated from data collected on Form EIA-914 (Monthly Natural Gas Production Report) for Federal Offshore Gulf of Mexico, Texas, Louisiana, New Mexico, Oklahoma, Texas, Wyoming, other states and lower 48 states. Alaska data are from the Alaska state government and included to obtain a U.S. total.

2015-01-01

295

ConocoPhillips Gas Hydrate Production Test  

SciTech Connect

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.

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

2013-06-30

296

Evaluation of six decontamination processes on actinide and fission product contamination  

SciTech Connect

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

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

1995-12-31

297

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

SciTech Connect

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.

Whitkop, P.G.

2001-06-26

298

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

NASA Astrophysics Data System (ADS)

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

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

2014-12-01

299

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

SciTech Connect

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.

Gilles Youinou; Andrea Alfonsi

2012-03-01

300

Comments on the possible roles of volatile fission products (cesium) in CABRI tests  

SciTech Connect

An investigation of information within the CABRI program that relates to the possible roles of volatile fission products (as represented by cesium) will be described. This study was partially motivated by the observation of localized /sup 137/Cs concentration peaks in the axial gamma scans of pins pre-irradiated to about 5% burnup (B.U.) level. In order to evaluate potential effects of such concentrations, a re-examination of the existing test data for the 1% B.U. pins was performed. A comparison of CABRI hodoscope fuel motion results and the pre-CABRI /sup 137/Cs axial concentration profiles revealed an approximate spatial correlation between the initial points of fuels dispersal and cesium concentration enhancements (seven of eight cases). 9 figs.

Lumpkin, A.H.

1986-01-01

301

Technical bases for estimating fission product behavior during LWR accidents. Technical report  

SciTech Connect

The objective of this report is to provide the Nuclear Regulatory Commission and the public with a description of the best technical information currently available for estimating the release of radioactive material during postulated reactor accidents, and to identify where gaps exist in our knowledge. This report focuses on those low probability-high consequence accidents involving severe damage to the reactor core and core meltdown that dominate the risk to the public. Furthermore, in this report particular emphasis is placed on the accident behavior of radioactive iodine, as (1) radioiodine is predicted to be a major contributor to public exposure, (2) current regulatory accident analysis procedures focus on iodine, and (3) several technical issues have been raised recently about the magnitude of iodine release. The generation, transport, and attenuation of aerosols were also investigated in some detail to assess their effect on fission product release estimates and to determine the performance of engineered safety features under accident conditions exceeding their design bases.

Not Available

1981-06-01

302

Electrochemistry of actinides and fission products in molten salts-Data review  

NASA Astrophysics Data System (ADS)

The thermodynamic and electrochemical properties of actinides and fission products in the molten salt determine the pyroprocessing separation performance. Extensive measurements have been carried out to provide fundamental data for evaluating the separation efficiency and technology feasibility of pyroprocessing although the technology has been very well developed in laboratory. The state of the art of fundamental data for substance or materials involved in pyropocessing will be reviewed in the present article. The available data will be summarized and reanalyzed. New correlations, which extend the available data to a broad range of applications, will be developed based on available data from different measurements. Further research topics on providing fundamental data that is needed for scaling the current laboratory technology to industrial applications are identified.

Zhang, Jinsuo

2014-04-01

303

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

SciTech Connect

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.

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

1991-04-01

304

Use of the linear accelerator for incinerating the fission products of /sup 137/Cs and /sup 90/Sr  

SciTech Connect

Transmutation of fission products /sup 137/Cs and /sup 90/Sr using the neutron produced by high energy proton collision with heavy nuclei were investigated. Because of the small thermal neutron cross section for (n,..gamma..) reaction of /sup 137/Cs (0.1 barn), a high neutron flux of 10/sup 17/ n/cm/sup 2/ sec is required to transmute /sup 137/Cs at a rate ten times faster than the natural decay. This range of high flux is attainable in the spallation reaction of high energy proton beam interact with liquid Pb target. The neutronic calculation by using NMTC, HIST3D, EPR, TAPEMAKER and ANISN codes indicates that the spallation neutron can transmute 222 kg /sup 137/Cs and 155 kg /sup 90/Sr fission products per year (at a rate of 10 and 30 times faster than their natural decay rate) by running a 300 mA, 1.5 GeV proton beam. Thus, if we transmute these fission products, just after a burning cycle, this accelerator can transmute these fission products produced in five or six 1000 MWe power plants.

Takahashi, H; Mizoo, N; Steinberg, M

1980-07-01

305

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

SciTech Connect

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

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

2009-01-01

306

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

NASA Astrophysics Data System (ADS)

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

Liu, X.-Y.; Uberuaga, B. P.; Nerikar, P.; Stanek, C. R.; Sickafus, K. E.

2010-10-01

307

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

E-print Network

ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Atomic Energy Agency, Vienna-A-1400, PO Box 100, Austria 8 National Institute of Standards and Technology Jozef Stefan Institute, Jamova 39, 1000 Ljubljana, Slovenia 11 Nuclear Research and Consultancy Group, P

Danon, Yaron

308

RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS  

SciTech Connect

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.

Clark, E.

2013-08-31

309

Dynamical theory of nuclear fission  

Microsoft Academic Search

The asymmetric fission problem is investigated from the dynamical theory point of view under a variety of conditions which cover most cases encountered in practice. Quantitative and qualitative results are obtained. In all the cases studied the dynamical theory cannot explain the experimental results of asymmetric mass distribution of fission products. NUCLEAR REACTION Fission. Asymmetric fission.

Peter Fong

1976-01-01

310

Nuclear fission  

Microsoft Academic Search

Nuclear fission is reviewed. The value of a channel analysis is noted in determining fission cross sections, fluctuations in fission widths and fission asymmetry. The importance of barrier height in any fission mode is emphasized. The difficulty of obtaining the mass distribution of fission fragments by a statistical theory is noted; this situation arises because the result is a strong

J. A. Wheeler

1956-01-01

311

Delayed beta- and gamma-ray production due to thermal-neutron fission of ²³⁹Pu: tabular and graphical spectral distributions for times after fission between 2 and 14000 sec  

Microsoft Academic Search

Fission-product decay energy-release rates were measured for thermal-neutron fission of ²³⁹Pu. Samples of mass 1 and 5 ..mu..g were irradiated for 1 to 100 s using the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were separately counted for times-after-fission between 2 and 14,000 s to yield spectral distributions N(E\\/sub ..gamma..\\/) vs E\\/sub

J. K. Dickens; T. R. England; T. A. Love; J. W. McConnell; J. F. Emergy; K. J. Northcutt; R. W. Peelle

1980-01-01

312

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

SciTech Connect

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

Gauntt, Randall O.

2010-04-01

313

Partitioning of fission products from irradiated nitride fuel using inductive vaporization  

SciTech Connect

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)

Shcherbina, N.; Kulik, D.A.; Kivel, N.; Potthast, H.D.; Guenther-Leopold, I. [Paul Scherrer Institut - PSI, Villigen 5232 (Switzerland)

2013-07-01

314

Modernizing the Fission Basis  

NASA Astrophysics Data System (ADS)

A recent Fission Product Review Panel study has identified important issues associated with the possible neutron energy dependence of the fission product isotope ^147Nd. As a result, we initiated a program at TUNL to obtain high-precision and self-consistent data for the energy dependence of fission product yields in the 1 to 15 MeV energy range. Three dual fission ionization chambers dedicated to ^235U, ^238U, and ^239Pu thick target foils and thin monitor foils, respectively, were exposed to neutron beams produced via the reactions ^2H(d,n)^3He and ^3H(d,n)^4He. After irradiation, the characteristic ? rays from specific fission products were recorded over a period of many weeks using HPGe detectors in a low-background environment. Results for the yield of seven fission isotopes obtained at 4.6, 9.0 and 14.8 MeV are reported.

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.; Wilhemy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

2012-10-01

315

Fission Xenon on Mars  

NASA Technical Reports Server (NTRS)

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.

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

2002-01-01

316

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

SciTech Connect

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.

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

1994-06-01

317

Yields of In and Sn products from thermal- and 14-MeV-neutron-induced fission of 235U  

NASA Astrophysics Data System (ADS)

The fractions of tin fission products formed independently and by decay of indium isotopes were determined for A=121, 123, 125, 127, and 128 from thermal-neutron-induced fission of 235U, and for A=121, 123, 127, and 128 from 14-MeV-neutron-induced fission of 235U. The procedure involved on-line chemical separation of indium and tin fission products by use of a continuous-extraction method during irradiation and beta-activity measurements of purified tin samples after decay of indium precursors. Measured fractions of tin isotopes formed directly or by beta decay combined with their cumulative yields and mass-number (chain) yields allowed calculation of independent and/or cumulative yields for the nuclides studied. The yields of indium isotopes agree with complementary technetium yields measured radiochemically, but they are lower than most indium yields measured mass spectrometrically and most complementary technetium yields measured with recoil separators. The yield data were used to determine mass- and charge-distribution parameters ?A', ?A', Y(Z), ?Z, and ?Z for thermal-neutron-induced fission of 235U. The values of ?A' and ?Z are similar to the averages of values derived for other elements and mass numbers. ?A' has a pronounced peak at Z=50, and the elemental yield decreases sharply from Y(Z=50)=4.0% to Y(Z=49)=0.1%. The ?Z function increases abruptly near Z=50, changing from ~-0.45 to positive values, as indicated by earlier experimental data, but inconsistent with the scission-point theoretical predictions. These effects, associated with the fifty-proton shell, diminish at higher excitation energy.

Semkow, Tomasz M.; Wahl, Arthur C.; Robinson, Larry

1984-12-01

318

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

NASA Astrophysics Data System (ADS)

The feasibility of directly irradiating five long-lived fission products (LLFPs: 79Se, 93Zr, 107Pd, 126Sn, and 135Cs, 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. 126Sn and 79Se 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 126Sn, 3.5 LWRs could be supported per year (at 75% beam availability). Furthermore, the addition of a 126Sn 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.

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

1995-09-01

319

Thermal Rate Constants of the NO2 Fission Reaction of Gas Phase r-HMX: A Direct ab Initio Dynamics Study  

E-print Network

Thermal Rate Constants of the NO2 Fission Reaction of Gas Phase r-HMX: A Direct ab Initio Dynamics was found along the reaction path. Microcanonical rate constants calculation shows the variational. The VT method was used for thermal rate constants calculation over a temperature range from 250 to 2000

Utah, University of

320

Numerical Simulations of Fission  

NASA Astrophysics Data System (ADS)

In this paper, we use the term fission to refer to the breakup of an equilibrium celestial body driven by rapid rotation. Historically, it was conjectured that fission would lead to splitting of a body directly into two or more pieces. Numerical hydrodynamic simulation techniques have now become sufficiently powerful to study the outcome of dynamic fission instabilities. We summarize recent work and present new simulations spanning a range of rotation rates and fluid compressibility. In the best resolved cases dynamic fission instability always leads to ejection of a ring or disk of debris rather thin one or a few discrete bodies. In this case, just as in most other lunar origin theories, a fission-product Moon must accrete out of a geocentric swarm of material. Intrinsic nonaxisymmetry of the remnant Earth after fission would prevent rapid recollapse of the swarm. The revised picture aleviates some of the problems associated with earlier versions of the fission theory. The two most serious remaining objections are that it is difficult to make the proto-Earth rotate fast enough to undergo fission and that the proto-Earth must be largely molten at the time it fissions. To overcome the first objection, it may be necessary to combine fission with the planetesimal impact theory. Some advantages of such a hybrid theory are discussed. The second objection cannot be fully assessed until more is known about the fission history and accretion of the proto-Earth.

Durisen, Richard H.; Gingold, Robert A.

1987-01-01

321

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

NASA Technical Reports Server (NTRS)

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

Kaiser, W. A.

1972-01-01

322

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

SciTech Connect

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.

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

1993-09-01

323

Effects of fission product incorporation on the microstructure of cubic zirconia  

NASA Astrophysics Data System (ADS)

Cesium, iodine and strontium ions have been implanted into yttria-stabilized cubic zirconia (YSZ) to determine the effects of fission product incorporation in YSZ that is considered as an inert nuclear fuel matrix. The ion implantation was conducted at room temperature to 110 21 ions/m 2 for each ion with ion energies ranging from 70 to 400 keV. The peak displacement damage level and the peak ion concentration in YSZ reached 205-330 displacement per atoms (dpa) and 11-26 at.%, respectively. The microstructure of the implanted YSZ was studied by in situ and cross-sectional transmission electron microscopy (TEM). In the iodine and strontium implanted samples, a damaged layer with a high density of defect clusters was observed, while in the cesium implanted specimen, most of the damaged layer is amorphous.Nanocrystalline precipitates were observed in the strontium implanted specimen after annealing at 1273 K. The results are discussed in terms of the ionic size, mobility and the solubility of the implanted species in YSZ.

Wang, L. M.; Wang, S. X.; Zhu, S.; Ewing, R. C.

2001-02-01

324

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

SciTech Connect

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.

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

2014-01-01

325

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

SciTech Connect

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

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

2004-10-01

326

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

NASA Astrophysics Data System (ADS)

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

Ugajin, M.; Itoh, A.

1994-10-01

327

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

SciTech Connect

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).

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

2012-01-01

328

Bio-gas production from alligator weeds  

NASA Technical Reports Server (NTRS)

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.

Latif, A.

1976-01-01

329

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

SciTech Connect

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

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

2004-03-29

330

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

SciTech Connect

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)

Ajlouni, Abdul-Wali M.S. [Ministry of Energy and Mineral Resources, Amman 11814 (Jordan)

2006-07-01

331

Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA  

Microsoft Academic Search

We report results of air monitoring started due to the recent natural catastrophe on 11 March 2011 in Japan and the severe ensuing damage to the Fukushima Dai-ichi nuclear reactor complex. On 1718 March 2011, we registered the first arrival of the airborne fission products 131I, 132I, 132Te, 134Cs, and 137Cs in Seattle, WA, USA, by identifying their characteristic gamma

J. Diaz Leon; D. A. Jaffe; J. Kaspar; A. Knecht; M. L. Miller; R. G. H. Robertson; A. G. Schubert

2011-01-01

332

Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA  

Microsoft Academic Search

We report results of air monitoring started due to the recent natural catastrophe on 11 March 2011 in Japan and the severe ensuing damage to the Fukushima Dai-ichi nuclear reactor complex. On 17-18 March 2011, we registered the first arrival of the airborne fission products 131-I, 132-I, 132-Te, 134-Cs, and 137-Cs in Seattle, WA, USA, by identifying their characteristic gamma

J. Diaz Leon; D. A. Jaffe; J. Kaspar; A. Knecht; M. L. Miller; R. G. H. Robertson; A. G. Schubert

2011-01-01

333

Interpretation of In-Pile Oscillation Experiments in the Minerve Facility for the Improvement of Fission Product Cross Sections  

Microsoft Academic Search

This document describes the interpretation of in-pile oscillation experiments in the Minerve facility at the CEA Cadarache. The objective of this study is the improvement of fission products (FPs) cross sections. The experimental device and the oscillation technique are described in a first part, then the interpretation method, based upon the deterministic Apollo2.8 code, is presented in a second part.

A. Gruel; P. Leconte; D. Bernard

2010-01-01

334

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

335

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

E-print Network

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 ...

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

2004-01-01

336

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

SciTech Connect

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.

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

2007-11-15

337

On the Cs,Te fission product-induced attack and embrittlement of stainless steel cladding in oxide fuel pins  

NASA Astrophysics Data System (ADS)

Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), attempts are made to rationalize the observed out-of-pile Cs: Te dependences of FCCI and FPLME incidence and severity, and their particular Cs, Te synergisms, in terms of Cs-Te thermochemistry and phase equilibria. Successful rationalization in the case of FPLME is taken to point up the critical importance of Te activity and Cs-Te physical state in the FPLME mechanism. A similar conclusion is reached for CCCT, the nonoxidative mode of FCCI, however oxidative modes of FCCI are concluded to rely more on the physical or catalytic properties of Cs-Te mixtures than on specific thermodynamic properties such as Te or Cs activities. The possibility of synergistic coupling between oxidative FCCI and FPLME in irradiated fuel pins is also examined, and it is concluded that although the available evidence does not support such coupling under monotonie loading, it is suggested as intergranular notch-sensitivity in FPLME under cyclic loading conditions.

Adamson, M. G.; Aitken, E. A.

1985-06-01

338

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

E-print Network

Synthesis Gas Production from Partial Oxidation of Methane with Air in AC Electric Gas Discharge K 73019 Received October 11, 2002 In this study, synthesis gas production in an AC electric gas discharge, Bangkok 10330, Thailand Lance L. Lobban and Richard G. Mallinson* Institute for Gas Utilization

Mallinson, Richard

339

Shale Gas Production: Potential versus Actual GHG Emissions  

E-print Network

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 ...

O'Sullivan, Francis

340

Natural gas production verification tests. Environmental assessment  

SciTech Connect

This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) in compliance with the requirements of the National Environmental Policy Act of 1969. The Department of Energy (DOE) proposes to fund, through a contract with Petroleum Consulting Services, Inc. of Canton, Ohio, the testing of the effectiveness of a non-water based hydraulic fracturing treatment to increase gas recovery from low-pressure, tight, fractured Devonian Shale formations. Although Devonian Shales are found in the Appalachian, Michigan, and Illinois Basins, testing will be done only in the dominant, historical five state area of established production. The objective of this proposed project is to assess the benefits of liquid carbon dioxide (CO{sub 2})/sand stimulations in the Devonian Shale. In addition, this project would evaluate the potential nondamaging (to the formation) properties of this unique fracturing treatment relative to the clogging or chocking of pores and fractures that act as gas flow paths to the wellbore in the target gas-producing zones of the formation. This liquid CO{sub 2}/sand fracturing process is water-free and is expected to facilitate gas well cleanup, reduce the time required for post-stimulation cleanup, and result in improved production levels in a much shorter time than is currently experienced.

Not Available

1992-02-01

341

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

SciTech Connect

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

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

1997-05-31

342

Identification of ?s isomers in the fission products of 241Pu(nth,f)  

NASA Astrophysics Data System (ADS)

Several ?s isomers were observed in neutron-rich nuclei in the mass range A=88-109. These nuclei are produced by thermal neutron-induced fission of 241Pu. The detection is based on time correlation between fission fragments selected by the LOHENGRIN spectrometer, at ILL (Grenoble) and the ?-ray or conversion electron emission of the isomers. Among a dozen isomers studied, three new ones have been observed. The decay schemes of 88mBr, 94mY, 96mRb, and 100mNb are discussed in the framework of the spherical shell model. The yields and isomeric ratios of all these isomers have been measured.

Genevey, J.; Ibrahim, F.; Pinston, J. A.; Faust, H.; Friedrichs, T.; Gross, M.; Oberstedt, S.

1999-01-01

343

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

SciTech Connect

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 5.21021 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.

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

2014-11-01

344

Measurement of product alignment in beam-gas chemiluminescent reactions  

E-print Network

Measurement of product alignment in beam-gas chemiluminescent reactions Michael G. Prisant, Charles of the chemiluminescent atom-diatom exchange reaction A + BC-.AB* + C under beam-gas conditions. The degree of product momentum appears as product rotational angular momentum. For beam-gas chemiluminescence, this implies

Zare, Richard N.

345

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

NASA Astrophysics Data System (ADS)

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--700C. 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.

Zhu, Sha

346

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

NASA Astrophysics Data System (ADS)

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.

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

347

Re-publication of the data from the BILL magnetic spectrometer: The cumulative $?$ spectra of the fission products of $^{235}$U, $^{239}$Pu, and $^{241}$Pu  

E-print Network

In the 1980s, measurements of the cumulative $\\beta$ spectra of the fission products following the thermal neutron induced fission of $^{235}$U, $^{239}$Pu, and $^{241}$Pu were performed at the magnetic spectrometer BILL at the ILL in Grenoble. This data was published in bins of 250 keV. In this paper, we re-publish the original data in a binning of 50 keV for $^{235}$U and 100 keV for $^{239}$Pu and $^{241}$Pu.

N. Haag; W. Gelletly; F. von Feilitzsch; L. Oberauer; W. Potzel; K. Schreckenbach; A. A. Sonzogni

2014-05-30

348

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

SciTech Connect

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.

Gohar, Y.

2000-11-15

349

Bio Gas Oil Production from Waste Lard  

PubMed Central

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

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

2011-01-01

350

Bio gas oil production from waste lard.  

PubMed

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

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

2011-01-01

351

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

SciTech Connect

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

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

2014-03-01

352

Ionic Liquid and Supercritical Fluid Hyphenated Techniques for Dissolution and Separation of Lanthanides, Actinides, and Fission Products  

SciTech Connect

This project is investigating techniques involving ionic liquids (IL) and supercritical (SC) fluids for dissolution and separation of lanthanides, actinides, and fission products. The research project consists of the following tasks: Study direct dissolution of lanthanide oxides, uranium dioxide and other actinide oxides in [bmin][Tf{sub 2}N] with TBP(HNO{sub 3}){sub 1.8}(H{sub 2}O){sub 0.6} and similar types of Lewis acid-Lewis base complexing agents; Measure distributions of dissolved metal species between the IL and the sc-CO{sub 2} phases under various temperature and pressure conditions; Investigate the chemistry of the dissolved metal species in both IL and sc-CO{sub 2} phases using spectroscopic and chemical methods; Evaluate potential applications of the new extraction techniques for nuclear waste management and for other projects. Supercritical carbon dioxide (sc-CO{sub 2}) and ionic liquids are considered green solvents for chemical reactions and separations. Above the critical point, CO{sub 2} has both gas- and liquid-like properties, making it capable of penetrating small pores of solids and dissolving organic compounds in the solid matrix. One application of sc-CO{sub 2} extraction technology is nuclear waste management. Ionic liquids are low-melting salts composed of an organic cation and an anion of various forms, with unique properties making them attractive replacements for the volatile organic solvents traditionally used in liquid-liquid extraction processes. One type of room temperature ionic liquid (RTIL) based on the 1-alkyl-3-methylimidazolium cation [bmin] with bis(trifluoromethylsulfonyl)imide anion [Tf{sub 2}N] is of particular interest for extraction of metal ions due to its water stability, relative low viscosity, high conductivity, and good electrochemical and thermal stability. Recent studies indicate that a coupled IL sc-CO{sub 2} extraction system can effectively transfer trivalent lanthanide and uranyl ions from nitric acid solutions. Advantages of this technique include operation at ambient temperature and pressure, selective extraction due to tunable sc-CO{sub 2} solvation strength, no IL loss during back-extraction, and no organic solvent introduced into the IL phase.

Wai, Chien M. [Univ. of Idaho, Moscow, ID (United States); Bruce Mincher

2012-12-01

353

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

SciTech Connect

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)

Harbachova, N.V.; Sharavarau, H.A. [Joint Institute of Power and Nuclear Research - 'Sosny' National Academy of Sciences, 99 Academic, A.K. Krasin Str., 220109 Minsk (Belarus)

2006-07-01

354

Authors' reply to Comment on the paper Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment  

NASA Astrophysics Data System (ADS)

We thank R. Konings et al. for their interest and their valuable critical discussion of our article regarding the fission product release from irradiated oxide fuel during thermal treatment and reply to their comments appearing in this issue. Their feedback stimulated us to give more details on the sampling procedure of investigated materials as well as the measurement procedure in order to exclude misunderstandings. The release curves for iodine and cesium are compared to blank profiles and reanalyzed to demonstrate the features of inductive heating approach applied in authors' recent study on FP release under inert and oxidizing conditions.

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

2014-04-01

355

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

Microsoft Academic Search

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

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

1978-01-01

356

Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA  

E-print Network

We report results of air monitoring started due to the recent natural catastrophe on March 11, 2011 in Japan and the severe ensuing damage to the Fukushima nuclear reactor complex. On March 17-18, 2011 we detected the first arrival of the airborne fission products 131-I, 132-I, 132-Te, 134-Cs, and 137-Cs in Seattle, WA, USA, by identifying their characteristic gamma rays using a germanium detector. The highest detected activity to date is <~32 mBq/m^3 of 131-I.

Leon, J Diaz; Knecht, A; Miller, M L; Robertson, R G H; Schubert, A G

2011-01-01

357

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

SciTech Connect

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.

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

2010-09-23

358

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

NASA Astrophysics Data System (ADS)

The present paper reports a study of FEMAXI-V code and related report on code benchmarking. Capabilities of the FEMAXI-V code to predict the thermal and fission gas release have been tested on MOX fuels in LWRs which has been done in SCK.CEN & Belgonucleaire by using PRIMO MOX rod BD8 irradiation experiment after V Sobolev as reported O. J. Ott [2]. Base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor (ramp power excursion). The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop, installed on a movable structure of the core periphery. The power variations were obtained by inwards/backwards movements of the loop in the core water. The preconditioning phase for rod BD8 occurred at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/ (cm.min), reaching a terminal peak power level of 395 W/cm that lasted for 20 hours.

Suwardi, Dewayatna, W.; Briyatmoko, B.

2012-06-01

359

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

SciTech Connect

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.

Slater, C.O.

1990-05-01

360

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

SciTech Connect

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.

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

2011-09-23

361

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

SciTech Connect

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 {sup 235,238}U and {sup 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 {sup 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 'ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,' Nuclear Data Sheets 107, 2931 (2006).

Chadwick, M.B.; Herman, M.; Author (s): Chadwick,M.B.; Herman,M.; Oblozinsky,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

362

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

NASA Astrophysics Data System (ADS)

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. Obloinsk, 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)].

Chadwick, M. B.; Herman, M.; Obloinsk, 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

363

Absorption of gamma-emitting fission products and activation products by rice under flooded and unflooded conditions from two tropical soils  

Microsoft Academic Search

Summary The absorption of gamma-emitting fission products106Ru,125Sb,137Cs and144Ce and activation products59Fe,58Co.54Mn and65Zn by rice plants grown on two contrasting tropical soils, namely, a blak soil (pellustert) and a laterite (oxisol), and the effects of flooding were studied under controlled conditions. Results indicated greater uptake of106Ru and125Sb from the black soil than from the laterite. In contrast, the uptake of144Ce and137Cs

T. J. D'Souza; K. B. Mistry

1980-01-01

364

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

NASA Astrophysics Data System (ADS)

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

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

2008-01-01

365

Volatile fission product behaviour during thermal annealing of irradiated UO 2 fuel oxidised up to U 3O 8  

NASA Astrophysics Data System (ADS)

The behaviour and release of fission products during high-temperature annealing of irradiated UO 2 samples have been studied as a function of the oxidation state. The behaviour of a sample pre-oxidised to U 3O 8 was compared to that of non-pre-treated fuel from the same pellet radial location. The Knudsen cell mass spectrometer technique was used up to 1900 K for the pre-oxidised sample and up to 2800 K for the untreated sample. Both types of tests were run in vacuum. The possible chemical forms of the different fission products in the bulk and in the vapour phase have been estimated from the release curves and microprobe analysis. This study concerns essentially iodine, tellurium, caesium, rubidium, strontium, barium, technetium and molybdenum, whose effusion behaviour was strongly affected by the pre-oxidation treatment, resulting in an almost complete release by 1900 K. Release of zirconium, the lanthanides and actinides was observed at temperatures >1900 K, reached only in the case of the non-pre-treated UO 2 experiments.

Hiernaut, J.-P.; Wiss, T.; Papaioannou, D.; Konings, R. J. M.; Rondinella, V. V.

2008-01-01

366

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

SciTech Connect

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.

Hikaru Hiruta; Gilles Youinou

2013-09-01

367

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

NASA Astrophysics Data System (ADS)

Soil samples have been taken in 2001 from the area of a 1951 release from an underground storage tank of 6.7 L of an aqueous solution of irradiated uranium (360 GBq). A simulation of the dispersion of the actinides and fission products was conducted in the laboratory using irradiated natural uranium, non-irradiated natural uranium and metal standards dissolved in acidic aqueous solutions and added to soil columns containing uncontaminated prairie soil. The lab soil columns were allowed 12 to 14 months for contaminant transport. Soil samples were analyzed using gamma-ray spectroscopy, neutron activation analysis (NAA) and liquid scintillation counting (LSC) to determine the elemental concentrations of U, Cs and Sr. Diffusion coefficients from the 50 year soil samples and the lab soil samples were determined. The measured diffusion coefficients from the field samples were 3.0 x 10-4 cm2 s-1 (Cs-137), 1.8 x 10-5 cm2 s-1 (U-238) and 2.6 x 10-3 cm2 s-1 (Sr-90) and the values determined from lab simulation were 5 x 10-6 cm 2 s-1 (Cs-137), 3 x 10-5 cm2 s-1 (U-238) and 1.9 x 10-5 cm 2 s-1 (Sr-90). The differences between the sets of diffusion coefficients can be attributed to differences in retardation effects, weather effects and changes in the soil characteristics when transporting, such as porosity. The analytical work showed that Cs-137 content of soil can be determined effectively using gamma-ray spectroscopy; U-238 content can be measured using NAA; and Sr-90 content can be measured using LSC. For non- and low-radioactive species, it was shown that both flame atomic absorption spectrometry (FAAS) and inductively-coupled plasma-mass spectrometry (ICP-MS) gave comparable results for Sr, Cs and Sm, with the average values ranging from 0.5 to 4.5 ppm of each other. The U-238 content results from NAA and from ICP-MS showed general agreement with an average difference of 81.3 ppm on samples having concentrations up to 988.2 ppm. The difference may have been due to matrix interference. It was determined through finite element modeling that 250 years after the 1951 release, the soil concentration of the three contaminant of U-238, Sr-90 and Cs-137 will be less than their respective soil clearance level values and therefore will not pose a long term environmental hazard. The fastest nuclide to reach the water table, at a depth of 45 m below the surface, at Suffield Site 27 was calculated to be Sr-90 after a period of 15,000 years. Therefore, it is not necessary to remove the subsurface soil at Site 27 for site decontamination but it is recommended that a "no-digging" policy, except for scientific research, be enforced at this site.

Sims, D. J.

368

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

SciTech Connect

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.

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

2007-12-01

369

Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly  

NASA Astrophysics Data System (ADS)

Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

da Cruz, D. F.; Rochman, D.; Koning, A. J.

2014-04-01

370

Measurement of fission products ? decay properties using a total absorption spectrometer  

NASA Astrophysics Data System (ADS)

In a nuclear reactor, the ? decay of fission fragments is at the origin of decay heat and antineutrino flux. These quantities are not well known while they are very important for reactor safety and for our understanding of neutrino physics. One reason for the discrepancies observed in the estimation of the decay heat and antineutrinos flux coming from reactors could be linked with the Pandemonium effect. New measurements have been performed at the JYFL facility of Jyvskyl with a Total Absorption Spectrometer (TAS) in order to circumvent this effect. An overview of the TAS technique and first results from the 2009 measurement campaign will be presented.

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

2013-12-01

371

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

Microsoft Academic Search

In this study, synthesis gas production in an AC electric gas discharge of methane and air mixtures at room temperature and ambient pressure was investigated. The objective of this work was to understand how the CH4\\/O2 feed mole ratio, ethane added, diluent gas, residence time, input power, applied frequency, and waveform, affected methane and oxygen conversions, product selectivities, and specific

K. Supat; A. Kruapong; S. Chavadej; Lance L. Lobban; Richard G. Mallinson

2001-01-01

372

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

SciTech Connect

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.

Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Dewayatna, Winter, E-mail: winter@batan.go.id [Center for Nuclear Fuel Technology, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Yos, E-mail: soj@batan.go.id [Center for Nuclear Equipment and Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

2014-09-30

373

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

PubMed Central

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

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

2014-01-01

374

Gas production strategy of underground coal gasification based on multiple gas sources.  

PubMed

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

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

2014-01-01

375

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

NASA Technical Reports Server (NTRS)

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.

Lustig, P. H.

1973-01-01

376

Standard test method for gamma energy emission from fission products in uranium hexafluoride and uranyl nitrate solution  

E-print Network

1.1 This test method covers the measurement of gamma energy emitted from fission products in uranium hexafluoride (UF6) and uranyl nitrate solution. It is intended to provide a method for demonstrating compliance with UF6 specifications C 787 and C 996 and uranyl nitrate specification C 788. 1.2 The lower limit of detection is 5000 MeV Bq/kg (MeV/kg per second) of uranium and is the square root of the sum of the squares of the individual reporting limits of the nuclides to be measured. The limit of detection was determined on a pure, aged natural uranium (ANU) solution. The value is dependent upon detector efficiency and background. 1.3 The nuclides to be measured are106Ru/ 106Rh, 103Ru,137Cs, 144Ce, 144Pr, 141Ce, 95Zr, 95Nb, and 125Sb. Other gamma energy-emitting fission nuclides present in the spectrum at detectable levels should be identified and quantified as required by the data quality objectives. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its us...

American Society for Testing and Materials. Philadelphia

2005-01-01

377

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

E-print Network

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.

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

378

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

PubMed Central

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

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

379

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

E-print Network

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.

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

2015-01-01

380

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

E-print Network

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.

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

2015-02-10

381

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

NASA Astrophysics Data System (ADS)

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

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

2014-11-01

382

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

PubMed

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(2)(-) 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(2)(-). 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. PMID:22005465

Kobashigawa, Shinko; Suzuki, Keiji; Yamashita, Shunichi

2011-11-01

383

Mass-yield distributions of fission products from photofission of 232Th induced by 45- and 80-MeV bremsstrahlung  

NASA Astrophysics Data System (ADS)

The mass-yield distributions of various fission products in the 45- and 80-MeV bremsstrahlung-induced fission of 232Th have been determined by using a recoil catcher and an offline ?-ray spectrometric technique in the electron linac at the Pohang Accelerator Laboratory, Korea. The mass-yield distributions were obtained from the fission-product yield data using charge-distribution corrections. The peak-to-valley (P/V) ratio, the average value of light mass () and heavy mass (), and the average number of neutrons () in the bremsstrahlung-induced fission of 232Th at different excitation energies were obtained from the mass-yield data. From the present measurements and the existing data from the 232Th(?,f) reaction and those from the 232Th(n,f) reaction at various energies, the following observations were obtained: (i) The mass-yield distributions in the 232Th(?,f) reaction at various energies are triple humped, similar to those of the 232Th(n,f) reaction. (ii) The yields of fission products for A = 133-134, A = 138-139, and A = 143-144 and their complementary products in the 232Th(?,f) reaction are higher than those of other fission products due to the nuclear structure effect. (iii) The yields of symmetric fission products for A = 133-134 and their complementary products in the 232Th(?,f) reaction are lower than those in the 232Th(n,f) reaction, whereas those for A = 143-144 and their complementary products are reversed. (iv) The result of increasing of the symmetric product yield causes the decreasing of the peak-to-valley ratio with increasing the excitation energy. However, it is surprising to see that the increasing trends for the symmetric products yields and the decreasing trends for the P/V ratio in the 232Th(?,f) and 232Th(n,f) reactions are not similar but those in the 238U(?,f) and 238U(n,f) reactions are similar to each other. (v) The average values of , , and at different excitation energies in the 232Th(?,f) and 232Th(n,f) reactions are similar but those in the 238U(?,f) and 238U(n,f) reactions are different.

Naik, H.; Goswami, A.; Kim, G. N.; Lee, M. W.; Kim, K. S.; Suryanarayana, S. V.; Kim, E. A.; Shin, S. G.; Cho, M.-H.

2012-11-01

384

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

SciTech Connect

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.

Scaglione, John M [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

385

Thermal reactor. [liquid silicon production from silane gas  

NASA Technical Reports Server (NTRS)

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.

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

1982-01-01

386

Accounting for Adsorbed gas and its effect on production bahavior of Shale Gas Reservoirs  

E-print Network

ACCOUNTING FOR ADSORBED GAS AND ITS EFFECT ON PRODUCTION BEHAVIOR OF SHALE GAS RESERVOIRS A Thesis by SALMAN AKRAM MENGAL Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment... of the requirements for the degree of MASTER OF SCIENCE August 2010 Major Subject: Petroleum Engineering ACCOUNTING FOR ADSORBED GAS AND ITS EFFECT ON PRODUCTION BEHAVIOR OF SHALE GAS RESERVOIRS A Thesis by SALMAN AKRAM MENGAL...

Mengal, Salman Akram

2010-10-12

387

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

NASA Astrophysics Data System (ADS)

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.

Bull, Geoffrey R.

388

Isomers in Fission Fragments  

SciTech Connect

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.

Urban, W.; Faust, H.; Jentschel, M.; Koester, U.; Krempel, J.; Materna, Th.; Mutti, P.; Soldner, T. [Institut Laue-Langevin, B.P. 156, F-38042 Grenoble Cedex 9 (France); Genevey, J.; Pinston, J. A.; Simpson, G. [Laboratoire de Physique Subatomique et de Cosmologie, IN2P3-CNRS/Universite J. Fourier, F-38026 Grenoble Cedex (France); Rzaca-Urban, T.; Zlomaniec, A.; Lukasiewicz, M. [Faculty of Physics, University of Warsaw, PL-00681 Warsaw (Poland); Sieja, K. [Gesellschaft fuer Schwerionenforschung, D-64291 Darmstadt (Germany); Nowacki, F.; Dorvaux, O.; Gall, B. J. P.; Roux, B. [Institut Pluridisciplinaire Hubert Curien, F-67037 Strasbourg Cedex (France); Dare, J. A. [Department of Physics and Astronomy, The University of Manchester, M13 9PL Manchester (United Kingdom)] (and others)

2009-01-28

389

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

SciTech Connect

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)

Ajlouni, Abdul-Wali M.S. [Ministry of Energy and Mineral Resources (Jordan)

2006-07-01

390

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

SciTech Connect

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.

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

2010-04-01

391

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

PubMed

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

Ko?l, Michal; vadlenkov, Marie; Mil?k, Jn; Rypar, Vojt?ch; Koleka, Michal

2014-07-01

392

Arrival time and magnitude of airborne fission products from the Fukushima, Japan, reactor incident as measured in Seattle, WA, USA  

E-print Network

We report results of air monitoring started due to the recent natural catastrophe on 11 March 2011 in Japan and the severe ensuing damage to the Fukushima Dai-ichi nuclear reactor complex. On 17-18 March 2011, we registered the first arrival of the airborne fission products 131-I, 132-I, 132-Te, 134-Cs, and 137-Cs in Seattle, WA, USA, by identifying their characteristic gamma rays using a germanium detector. We measured the evolution of the activities over a period of 23 days at the end of which the activities had mostly fallen below our detection limit. The highest detected activity amounted to 4.4 +/- 1.3 mBq/m^3 of 131-I on 19-20 March.

J. Diaz Leon; D. A. Jaffe; J. Kaspar; A. Knecht; M. L. Miller; R. G. H. Robertson; A. G. Schubert

2011-08-23

393

GASCAP: Wellhead Gas Productive Capacity Model documentation, June 1993  

SciTech Connect

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.

Not Available

1993-07-01

394

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

SciTech Connect

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.

Henager, Charles H.; Jiang, Weilin

2014-11-01

395

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

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.

S. Frank

2009-09-01

396

Production of biodiesel using expanded gas solvents  

SciTech Connect

A method of producing an alkyl ester. The method comprises providing an alcohol and a triglyceride or fatty acid. An expanding gas is dissolved into the alcohol to form a gas expanded solvent. The alcohol is reacted with the triglyceride or fatty acid in a single phase to produce the alkyl ester. The expanding gas may be a nonpolar expanding gas, such as carbon dioxide, methane, ethane, propane, butane, pentane, ethylene, propylene, butylene, pentene, isomers thereof, and mixtures thereof, which is dissolved into the alcohol. The gas expanded solvent may be maintained at a temperature below, at, or above a critical temperature of the expanding gas and at a pressure below, at, or above a critical pressure of the expanding gas.

Ginosar, Daniel M [Idaho Falls, ID; Fox, Robert V [Idaho Falls, ID; Petkovic, Lucia M [Idaho Falls, ID

2009-04-07

397

Nuclear fission  

Microsoft Academic Search

The object of the book is to provide a comprehensive account of present ; understanding of nuclear fission. While it is written at an introductory level ; for students of the physics and chemistry of fission, it also attempts to cover ; recent developments at a sufficient depth to make the volume valuable to research ; scientists. The theoretical framework

R. Vandenbosch; J. R. Huizenga

1973-01-01

398

Tempest gas turbine extends EGT product line  

Microsoft Academic Search

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

Chellini

1995-01-01

399

Mobile neutron/gamma waste assay system for characterization of waste containing transuranics, uranium, and fission/activation products  

SciTech Connect

A new integrated neutron/gamma assay system has been built for measuring 55-gallon drums at Pacific Northwest Laboratory. The system is unique because it allows simultaneous measurement of neutrons and gamma-rays. This technique also allows measurement of transuranics (TRU), uranium, and fission/activation products, screening for shielded Special Nuclear Material prior to disposal, and critically determinations prior to transportation. The new system is positioned on a platform with rollers and installed inside a trailer or large van to allow transportation of the system to the waste site instead of movement of the drums to the scanner. The ability to move the system to the waste drums is particularly useful for drum retrieval programs common to all DOE sites and minimizes transportation problems on the site. For longer campaigns, the system can be moved into a facility. The mobile system consists of two separate subsystems: a passive Segmented Gamma Scanner (SGS) and a {open_quotes}clam-shell{close_quotes} passive neutron counter. The SGS with high purity germanium detector and {sup 75}Se transmission source simultaneously scan the height of the drum allowing identification of unshieled {open_quotes}hot spots{close_quotes} in the drum or segments where the matrix is too dense for the transmission source to penetrate. Dense segments can flag shielding material that could be used to hide plutonium or uranium during the gamma analysis. The passive nuetron counter with JSR-12N Neutron Coincidence Analyzer measures the coincident neutrons from the spontaneous fission of even isotopes of plutonium. Because high-density shielding produces minimal absorption of neutrons, compared to gamma rays, the passive neutron portion of the system can detect shielded SNM. Measurements to evaluate the performance of the system are still underway at Pacific Northwest Laboratory.

Davidson, D.R. [Canberra Industries, Inc., Meriden, CT (United States); Haggard, D.; Lemons, C. [Pacific Northwest Lab., Richland, WA (United States)

1994-12-31

400

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

...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...

2012-07-01

401

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

DOEpatents

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.

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

1995-01-01

402

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

NASA Astrophysics Data System (ADS)

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.

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

403

Measuring micro-organism gas production  

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

404

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

SciTech Connect

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.

McGuire, P.L.

1981-01-01

405

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

SciTech Connect

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.

Kobashigawa, Shinko, E-mail: kobashin@nagasaki-u.ac.jp [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)] [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan); Suzuki, Keiji; Yamashita, Shunichi [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)] [Atomic Bomb Disease Institute, Course of Life Sciences and Radiation Research, Nagasaki University Graduate School of Biomedical Sciences, 1-12-4 Sakamoto, Nagasaki 852-8523 (Japan)

2011-11-04

406

78 FR 33051 - Notification of Proposed Production Activity, The Gas Company, LLC dba Hawai'i Gas, Subzone 9F...  

Federal Register 2010, 2011, 2012, 2013, 2014

...authorized by the FTZ Board. Production under FTZ procedures could...components used in export production. On its domestic sales...natural gas, carbon dioxide, hydrogen, hydrocarbon gas mixtures...reduced on foreign status production equipment. The...

2013-06-03

407

Interference effects in nuclear fission  

Microsoft Academic Search

Various interference effects governing the character of angular distributions of binary and ternary nuclear fission products\\u000a and P-odd, P-even, and T-odd asymmetries in these angular distributions have been studied within the quantum theory of spontaneous and low-energy\\u000a induced nuclear fission.

S. G. Kadmensky; L. V. Titova

2007-01-01

408

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions  

SciTech Connect

The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias and uncertainty results based on a quality-assurance-controlled prerelease version of the Scale 6.1 code package and the ENDF/B-VII nuclear cross section data.

Radulescu, Georgeta [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

409

Local and Medium Range Order Around Fission Products in Inactive Waste Glasses: Implication for Glass Structure and Stability  

NASA Astrophysics Data System (ADS)

Borosilicate glasses are used to store high level nuclear waste in France (R7T7 glass). The structure of the glass around elements such as fission products controls important parameters as the homogeneity of the glass and/or the melted glass rheology. Data on the local and medium range order structure of these glasses could help improving the resistance toward leaching and/or irradiation, in relation with surface or geological storage of these vitrified wastes. Due to the complex composition of these glasses (up to 30 oxides), chemically selective methods are required to understand the environment of elements. X-ray Absorption Spectroscopy (XAS) is, from this point of view, a powerful tool as it provides a direct access to the investigation of the structure around specific cations in this multicomponent amorphous material, to specify their role in the glass durability. We will present different XAS studies (synchrotrons in LURE and ESRF, France) on the inactive amorphous analog for the R7T7 glass (the SON 68 glass). This report will illustrate the potentialities of this approach through the determination of the environment around fission products such as Zr, Zn and Mo. XAS shows the peculiarity of the sites occupied by these glass components of technological interest. Coordination numbers are shown to be systematically smaller than in crystalline compounds with close composition. Below the definition of the sites occupied by the chemical elements, XAS allows to detect some degree of medium range order which gives insight on the bonding of the site to the poymeric borosilicate network and allow to link precisely experimental data to theoretical calculations. Eventually, XAS is used to study the interaction between noble metals (Pd and Ru) and the glassy matrix. These elements are at the origin of small precipitates that induce changes in the melt vicosity. They occur as a result of the non-insertion of these elements in the glassy matrix. To accurate and precise structural interpretations, a direct comparison with MD calculations on simplified nuclear glass comprising 5 oxides, is performed.

Galoisy, L.; Calas, G.; Ghaleb, D.; Morin, G.

2002-12-01

410

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

Code of Federal Regulations, 2013 CFR

...2013-04-01 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...

2013-04-01

411

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

Code of Federal Regulations, 2014 CFR

...2014-04-01 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...

2014-04-01

412

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

Code of Federal Regulations, 2012 CFR

...2012-04-01 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...

2012-04-01

413

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

Code of Federal Regulations, 2011 CFR

...2011-04-01 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...

2011-04-01

414

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

SciTech Connect

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.

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

415

Low permeability gas reservoir production using large hydraulic fractures  

E-print Network

LOVT PERMEABILITY GAS RESERVOIR PRODUCTION USING LARGE HYDRAULIC FRACTURES A Thesis by STEPHEN ALLEN HOLDITCH Approved as to style and content by: ( airman of Committee) (Head of Department) (Me er) (Member) (Membe r) (Member) (Member...) August 1970 111 ABSTRACT Low Permeability Gas Reservoir Production Using Large Hydraulic Fractures. (August 1970) Stephen Allen Holditch, B. S. , Texas ARM University Directed by: Dr, R. A. Morse There has been relatively little work published...

Holditch, Stephen A

1970-01-01

416

Mitigating Accidents In Oil And Gas Production Facilities  

NASA Astrophysics Data System (ADS)

Integrated operations are increasingly used in oil and gas production facilities to improve yields, reduce costs and maximize profits. They leverage information and communications technology (ICT) to facilitate collaboration between experts at widely dispersed locations. This paper discusses the safety and security consequences of implementing integrated operations for oil and gas production. It examines the increased accident risk arising from the tight coupling of complex ICT and SCADA systems, and proposes technological, organizational and human factors based strategies for mitigating the risk.

Johnsen, Stig

417

Process for production desulfurized of synthesis gas  

DOEpatents

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.

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

1993-01-01

418

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

SciTech Connect

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.

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

419

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

USGS Publications Warehouse

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.

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

420

Ground-state ?--branching intensities of several fission-product isotopes measured using a total absorption ?-ray spectrometer  

NASA Astrophysics Data System (ADS)

The final set of results of "ground-state" ?--branching intensities obtained in a program of systematic study of those regions of the fission-product nuclides accessible to investigation using the 252Cf-based INEL ISOL facility are presented. A total absorption ?-ray spectrometer, operating in a 4??-? coincidence mode, was used to obtain these "ground-state" ?--branching intensities; where here the "ground-state" is defined to include all states below a selected ?-ray discriminator level. Results obtained for 89Rb, 90gRb, 91Rb, 93Rb, 93Sr, 94Sr, 94Y, 95Sr, 95Y, 140Cs, 142La, 143Ba, 143La, 144Ba, 144La, 145Ba, 145La, 146Ce, 146Pr, 147Ce, 147Pr, 148Ce, 148Pr, (2.27 min), 149Pr, 149Nd, 151Pr, 151Nd, 152Pm (4.1 min), 153Nd, 155Nd, 157Pm, 157Sm, 158Sm and 158Eu are presented and compared with existin