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1

Transport time of volatile and nonvolatile fission products in a gas jet  

Microsoft Academic Search

Transport times for volatile and nonvolatile fission products in a gas jet were determined using the facility at the Ford Nuclear Reactor of the University of Michigan. A mixture of ethylene and nitrogen was used to sweep the fission products from the target chamber in the gas jet. Activated charcoal traps [C] and quartz wool traps [QW] were used to

N. Davis; E. T. Contis; K. Rengan; H. C. Griffin

1994-01-01

2

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

Microsoft Academic Search

An off-gas treatment system is being developed for the collection of volatile fission products during a head-end processing step. The head-end processing step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Three volatile fission products have been targeted for trapping on distinct zones of the

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

2008-01-01

3

Transport time of volatile and nonvolatile fission products in a gas jet  

NASA Astrophysics Data System (ADS)

Transport times for volatile and nonvolatile fission products in a gas jet were determined using the facility at the Ford Nuclear Reactor of the University of Michigan. A mixture of ethylene and nitrogen was used to sweep the fission products from the target chamber in the gas jet. Activated charcoal traps [C] and quartz wool traps [QW] were used to collect the volatile and nonvolatile fission products respectively. The trap was positioned in front of a HPGe detector. A "stopped-flow" technique was used for the transport time measurement. The gas flow was controlled with electrically operated valves; the application of power to the valves also triggered the counting in multiscaler mode. Measurements were carried out for two target pressures. For each pressure a number of measurements were done with the charcoal and the quartz wool traps. For a target pressure of 4 psi above atmosphere transport times of 432 41 and 432 23 ms were obtained for the volatile [C] and nonvolatile [QW] fission products respectively; at about atmospheric pressure the corresponding values were 458 33 and 443 38 ms. The values indicated that there is no significant difference in the transport time for the volatile and nonvolatile fission products in a gas jet.

Davis, N.; Contis, E. T.; Rengan, K.; Griffin, H. C.

1994-12-01

4

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

Microsoft Academic Search

A helium gas-jet system for the transport of fission products from the research reactor TRIGA Mainz has been developed, characterized and tested within the TRIGA-SPEC experiment. For the first time at TRIGA Mainz carbon aerosol particles have been used for the transport of radionuclides from a target chamber with high efficiency. The radionuclides have been identified by means of ?-spectroscopy.

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

2010-01-01

5

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

6

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

NASA Astrophysics Data System (ADS)

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

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

2010-02-01

7

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

8

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

9

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

10

Fission products experimental programme  

SciTech Connect

The 'Fission Products (FPs)' experimental programme was carried out from 1998 to 2004 in CEA/Valduc Apparatus B in the framework of the IRSN-AREVA NC Common Interest Program dealing with 'burnup credit' studies. It aims at compensating for the lack of critical experiments with FPs in the literature and validating a selection of 6 stable, non-volatile, fission products representing half of the irradiated fuel absorption of all fission products: {sup 103}Rh, {sup 133}Cs, {sup 143}Nd, {sup 149}Sm, {sup 152}Sm and {sup 155}Gd. This paper provides a more exhaustive approach of the k{sub eff} results, uncertainties and tendencies associated with the different steps of the 'Fission Products' programme. (authors)

Leclaire, N. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Anno, J. [Commissariat a l'Energie Atomique CEA, Institut de Protection et de Surete Nucleaire (CEA/IPSN) (France); Girault, E. [Commissariat a l'Energie Atomique CEA, Centre de Valduc, 21120 Is-Sur-Tille (France); Letang, E. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay-aux-Roses Cedex (France)

2006-07-01

11

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

12

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

13

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

14

Gas-leaking fuel elements number and fission gas product coolant volumetric activities assessment in the VVER-440 nuclear power plant  

NASA Astrophysics Data System (ADS)

In a nuclear power plant it is required to monitor continuously the number of gas-leaking fuel elements and the contamination level of the primary coolant by fission gas products. It is proposed to use the radiation monitoring system equipped with the computer technics provided with the suitable program package for fulfilment this requirements. The input data to start up the program consists of the 88Kr volumetric activity measured by the radiation monitoring system and three actual technological parameters: coolant temperature at inlet, thermal power and coolant flow rate.

Szuta, Marcin

1992-07-01

15

Downstream behavior of fission products  

SciTech Connect

The downstream behavior of fission products has been investigated by injecting mixtures of CsOH, CsI, and Te into a flowing steam/hydrogen stream and determining the physical and chemical changes that took place as the gaseous mixture flowed down a reaction duct on which a temperature gradient (1000/sup 0/ to 200/sup 0/C) had been imposed. Deposition on the wall of the duct occurred by vapor condensation in the higher temperature regions and by aerosol deposition in the remainder of the duct. Reactions in the gas stream between CsOH and CsI and between CsOH and Te had an effect on the vapor condensation. The aerosol was characterized by the use of impingement tabs placed in the gas stream.

Johnson, I.; Farahat, M.K.; Settle, J.L.; Johnson, C.E.; Ritzman, R.

1986-01-01

16

FISSION GAS YIELD IN URANIUM  

Microsoft Academic Search

The yield of xenon and krypton in irradiated uranium was found to vary ; with neutron flux from 4.8 cm³(STP)\\/cm³ uranium-atom percent burnup ; at 10¹² neut\\/cm²-sec to neut\\/cm²-sec as a result of neutron ; capture by the unstable isotopes Xe¹³³ and Xe¹³⁵. These results are ; applicable only to uranium irradiated in a thermal flux The fission gas yields

1962-01-01

17

Study of prediction of fission product behaviors in severe accident. Scoping experiment and basic findings of Cs, Ba and Sr.  

National Technical Information Service (NTIS)

Fission product release experiment (VEGA: Verification Experiments for fission product Gas/Aerosol release) has been started at JAERI, aiming at accurate source term evaluation during severe accident. Behaviors of fission products released from fuels larg...

M. Yamawaki J. Huang M. Tonegawa F. Ono M. Yasumoto

1998-01-01

18

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

SciTech Connect

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

Rider, B.F.

1995-05-01

19

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

NASA Astrophysics Data System (ADS)

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

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

1996-10-01

20

Short-lived fission products as a diagnostics tool for studying atom and ion behavior in a gas-based laser ion source  

NASA Astrophysics Data System (ADS)

A striking difference between gamma spectra of neutron-rich Rh isotopes obtained with the laser ion source at the Leuven Isotope Separator On-Line facility has been observed depending on the mode of operation. Although the global production rate of 112Rh g,m nuclei decreases considerably, the ratio between the productions of 112Rh g and 112Rh m increases strongly when no laser ionisation is used. This effect is caused by 112Ru atoms which decay during gas evacuation of the ion source gas cell, thereby producing 112Rh g ions. The comparison in time behaviour of reaction produced ions, ?-decay produced ions and laser produced ions makes it possible to study and characterise the different processes in the gas cell. The influence of these processes has to be considered when extracting nuclear information such as the relative feeding of different short-lived isomers and isotopes and fission cross-sections in a particular mass chain.

Weissman, L.; Prasad, N. V. S. V.; Bruyneel, B.; Huyse, M.; Kruglov, K.; Kudryavtsev, Y.; Muller, W. F.; Van Duppen, P.; Van Roosbroeck, J.

2002-05-01

21

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

22

Antiproton Powered Gas Core Fission Rocket  

NASA Astrophysics Data System (ADS)

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

2005-02-01

23

Antiproton Powered Gas Core Fission Rocket  

NASA Astrophysics Data System (ADS)

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, T.

24

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

25

ORNL fission product release tests VI-6.  

National Technical Information Service (NTIS)

The ORNL fission product release tests investigate release and transport of the major fission products from high-burnup fuel under LWR accident conditions. The two most recent tests (VI-4 and VI-5) were conducted in hydrogen. In three previous tests in th...

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

1991-01-01

26

Calculations on fission gas behaviour in the high burnup structure  

Microsoft Academic Search

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing

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

2006-01-01

27

Fuel and fission product release from sodium  

SciTech Connect

The NALA program at Kernforschungszentrum Karlsruhe is concerned with the release of fuel and fission products from hot or boiling sodium pools (radiological secondary source term) in a liquid-metal fast breeder reactor accident scenario with tank failure. The main concern is to determine retention factors (RF), to uncover the most essential parameters that influence the RF values, and to describe the way they do it. In the framework of the last NALA series, NALA IIIc, the influence of sodium-concrete interaction was investigated, partly with subsequent sodium burning. In our experiments, [approx]3 kg of sodium and added pieces of concrete reaching from 4 to 40 g was used. The composition of the concrete was suitable for shielding and construction as used in the SNR-300 reactor. Fuel was simulated by 20-[mu]m particles of depleted UO[sub 2], and CeO[sub 2], NaI, and TeO[sub 2] were used as fission products. Most experiments were performed in an inert argon gas atmosphere with monitored hydrogen development. In some cases, the preheated pool was allowed to come into contact with ambient air, which caused an ordinary sodium fire. For the latter case, we used the 220-m[sup 3] FAUNA vessel as an outer containment and collected the fire aerosols by a trap and subsequent filters for analysis.

Sauter, H. (Kernforschungszentrum Karlsruhe (Germany))

1992-01-01

28

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

NASA Technical Reports Server (NTRS)

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

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

1967-01-01

29

Actual Point About Fission Products Vitrification.  

National Technical Information Service (NTIS)

The main characteristics concerning the continuous vitrification process for the confinement of fission product solutions operated at AVM are summarized. The general principle of a vitrification plant is described. The AVM plant efficiency as also its con...

R. Bonniaud

1982-01-01

30

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

31

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

32

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

33

DISTRIBUTION OF FISSION PRODUCT CONTAMINATION IN THE SRE  

Microsoft Academic Search

BS>In the safety analysis of sodium-cooled reactors, a remaining area of ; significant uncertainty was the fate of various fission products that may be ; released to the coolant in the event of a fuel element failure; that is, the ; degree of their retainment in the sodium, movement to the reactor cover gas ; system or deposition in other

R. S. HART

1962-01-01

34

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

35

Exposure Rates from Experimentally Fractionated Fission Products.  

National Technical Information Service (NTIS)

Exposure rates, at three feet above a uniformly contaminated smooth plane, were calculated from recent experimentally measured gamma-ray spectra of fractionated products of thermal-neutron fission of 235U. The fractionation had been carried out by very ea...

L. R. Bunney D. Sam

1971-01-01

36

PRODUCTION OF VOID AND PRESSURE BY FISSION TRACK NUCLEATION OF RADIOLYTIC GAS BUBBLES DURING POWER BURSTS IN A SOLUTION REACTOR  

Microsoft Academic Search

The Kinetic Experiment on Water Boiler (KEWB) reactor is a 50-kw aqueous ;\\u000a homogeneous research reactor which was designed to study the safety ;\\u000a characteristics and dynamic behavior of this class of reactors. When the reactor ;\\u000a is placed on a short-period power transient, its aqueous uranyl sulfate fuel ;\\u000a solution becomes rapidly supersaturated with H gas produced by the

P. Spiegler; C. F. Jr. Bumpus; A. Norman

1962-01-01

37

Fission product release in high-burn-up UO 2 oxidized to U 3O 8  

NASA Astrophysics Data System (ADS)

Results of oxidation experiments on high-burn-up UO 2 are presented where fission-product vaporisation and release rates have been measured by on-line mass spectrometry as a function of time/temperature during thermal annealing treatments in a Knudsen cell under controlled oxygen atmosphere. Fractional release curves of fission gas and other less volatile fission products in the temperature range 800-2000 K were obtained from BWR fuel samples of 65 GWd t -1 burn-up and oxidized to U 3O 8 at low temperature. The diffusion enthalpy of gaseous fission products and helium in different structures of U 3O 8 was determined.

Colle, J. Y.; Hiernaut, J.-P.; Papaioannou, D.; Ronchi, C.; Sasahara, A.

2006-02-01

38

Nondestructive fission gas release measurement and analysis  

SciTech Connect

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

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

1993-01-01

39

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

40

Distribution of Independent Fission-Product Yields to Isomeric States.  

National Technical Information Service (NTIS)

A simple one-parameter model is presented for calculating the distribution of independent yield strength between ground and isomeric states of primary fission products formed by neutron-induced fission of actinide nuclei. Yield branching ratios are calcul...

D. G. Madland T. R. England

1976-01-01

41

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

42

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

SciTech Connect

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

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

2005-11-01

43

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

44

Fission Product Iodine Carry-over from Experimental Apparatus.  

National Technical Information Service (NTIS)

Measurements of fission product iodine carry-over have been made on a small scale laboratory test unit at one atmosphere. Fission product iodine in the water having several chemical species such as I exp 0 , I exp - , I exp +5 , and organic, the carry-ove...

H. Tone S. Okagawa

1977-01-01

45

(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

46

Heterogeneous nucleation of fission gas bubbles and gas migration in uranium dioxide  

Microsoft Academic Search

The number and size of fission gas bubbles precipitated in irradiated uranium dioxide are calculated from a theory based on a balance between nucleation of bubbles at vacancy clusters produced by fission fragments and the agglomeration of bubbles by random motion. The distribution of bubble sizes is determined by gas atom capture, bubble agglomeration and irradiation re-solution. Irradiation re-solution exceeds

A. D. Whapham

1972-01-01

47

Interaction of noble-metal fission products with pyrolytic silicon carbide  

SciTech Connect

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO/sub 2/ or UC/sub 2/ are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group.

Lauf, R.J.; Braski, D.N.

1982-01-01

48

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

SciTech Connect

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

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

2006-10-01

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

Yields of short-lived fission products following fast fission of U-238  

NASA Astrophysics Data System (ADS)

Fission-product yields following neutron-induced fission of 238U have been measured at the UMass Lowell 1-MW research reactor using gamma-ray spectroscopy. High- resolution gamma-ray spectra of aggregate fission products have been measured at 13 delay-time intervals ranging from 0.3 s to 4,000s after fission. A helium-jet system was used to rapidly transfer fission products from the fission chamber to a low-background counting area where they were sprayed onto a moving tape. The tape, whose speed determined the observed delay time, carried the products to a high-purity germanium detector. The use of beta-gamma coincidence reduced the background by about two orders of magnitude, and further improvement in the peak-to-background ratio was obtained by using a NaI(Tl) annulus for Compton suppression. The helium-jet system has been shown to give uniform transfer of fission products over the mass range studied. Cumulative and independent yields of fission products are calculated from the time-evolution of relative line intensities extracted from the aggregate spectra. An average of four lines per nuclide were used in this determination. Only lines showing the correct time evolution and relative intensity were used to assure there was no contamination from lines with similar energies. Measured nuclide yields are compared to those given in the ENDF/B-VI evaluation. Yields for 63 nuclides were determined, including twenty- one nuclides with halflives less than 2 s. Eleven determinations of yields were made for nuclides with isomeric states.

Campbell, Joann Marie

1997-07-01

51

Recent MELCOR and VICTORIA Fission Product Research at the NRC  

Microsoft Academic Search

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission-

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

1999-01-01

52

JNDC nuclear data library of fission products. Second version.  

National Technical Information Service (NTIS)

The second version of the JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library is described in this report. The library contains nuclear decay and fission yield data for 1078 unstable and 149 stable FP nuclides, and neutron cro...

K. Tasaka J. Katakura H. Ihara T. Nakagawa H. Takano

1990-01-01

53

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

NASA Astrophysics Data System (ADS)

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

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

2005-12-01

54

Thermodynamic treatment of noble metal fission products in nuclear fuel  

NASA Astrophysics Data System (ADS)

Based on a critical evaluation of the literature, a comprehensive thermodynamic model has been developed for the complete quinary system involving the noble metal fission products in nuclear fuel: Mo-Pd-Rh-Ru-Tc. This treatment was based on the foundation of ten binary systems and an interpolation scheme. The thermodynamic model has been demonstrated to fit the available experimental data for the ternary sub-systems. This work can be used with other models for potentially non-stoichiometric UO 2+ x containing fission products, as well as data for other phases, to assess the chemical form of fission products in irradiated fuel material.

Kaye, M. H.; Lewis, B. J.; Thompson, W. T.

2007-06-01

55

Bulk and surface controlled diffusion of fission gas atoms  

SciTech Connect

Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. [1-6]. One of the most commonly used models in fuel performance codes was published by Massih and Forsberg [3,4,6]. This model is similar to the early Booth model [1] in that it applies an equivalent sphere to separate bulk UO{sub 2} from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D{sub 3}) and intermediate (D{sub 2}) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's [8-10] and Matzke's data [12]. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO{sub 2{+-}x}, which compare favorably to available experiments. This is an extension of previous work [13]. In particular, it applies improved chemistry models for the UO{sub 2{+-}x} nonstoichiometry and its impact on the fission gas activation energies. The derivation of these models follows the approach that used in our recent study of uranium vacancy diffusion in UO{sub 2} [14]. Also, based on the calculated DFT data we analyze vacancy enhanced diffusion mechanisms in the intermediate temperature regime. In addition to vacancy enhanced diffusion we investigate species transport on the (111) UO{sub 2} surface. This is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO{sub 2} under irradiation, for which surface diffusion could be the rate-limiting transport step. Diffusion of such bubbles constitutes an alternative mechanism for mass transport in these materials.

Andersson, Anders D. [Los Alamos National Laboratory

2012-08-09

56

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

57

Fission product behavior distribution in the TMI-2 reactor building.  

National Technical Information Service (NTIS)

This paper summarizes the results of the examinations performed on samples from the reactor building surfaces and water and samples obtained from the reactor coolant system. The total quantities of fission products, fuel, and core material elements measur...

C. V. McIsaac R. Kohli R. S. Denning D. W. Akers

1988-01-01

58

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

59

Release of fission products from silicide fuel at elevated temperatures  

Microsoft Academic Search

An out-of-pile experiment on the release of fission products, such as [sup 131]I, [sup 133]Xe, [sup 85]Kr, [sup 137]Cs, and [sup 103]Ru, from silicide fuel was carried out to study fission-product retention capability for the safety evaluation of the Japan Materials Testing Reactor (JMTR) core conversion. The UAl[sub x]-Al fuel was included in the experiment to be compared with silicide

Y. Futamura; M. Saito; R. Oyamada; F. Sakurai; Y. Komori; J. Saito; T. Iwai; M. Shimizu; T. Nakagawa

2009-01-01

60

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

SciTech Connect

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

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

1987-01-01

61

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

Microsoft Academic Search

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

1980-01-01

62

Fission product release from ZrC-coated fuel particles during post-irradiation heating at 1800 and 2000C  

Microsoft Academic Search

The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particles for high-temperature gas-cooled reactors. Post-irradiation heating tests of the ZrC-Triso coated UO2 particles were performed at 1800C for 3000 h and at 2000C for 100 h to study the release behavior of fission products. The fission gas release monitoring and the X-ray

Kazuo Minato; Toru Ogawa; Kousaku Fukuda; Hajime Sekino; Isamu Kitagawa; Naoaki Mita

1997-01-01

63

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

64

Isomeric yield ratios of fission products in proton-induced fission of 232 Th  

Microsoft Academic Search

Isomeric yield ratios of 11 fission products were measured in the system of 13 MeV proton-induced fission of232Th by an on-line ion-guide isotope separator. It was found that the closed shell structures of primary fragments and their\\u000a complementary fragments affect the isomeric yield ratios. Isomeric yield ratios of121Cd (11\\/2?, 3\\/2+) and135Xe (11\\/2?, 3\\/2+) were measured precisely in the proton energy

S. Goto; D. Kaji; H. Kudo; M. Fujita; T. Shinozuka; M. Fujioka

1999-01-01

65

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

66

Steady-state and transient fission gas release and swelling model for LIFE-4. [LMFBR  

SciTech Connect

The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability.

Villalobos, A.; Liu, Y.Y.; Rest, J.

1984-06-01

67

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

68

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

69

Fission product release from fuel under severe accident conditions  

SciTech Connect

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

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

1993-03-01

70

A review of fission product sorption in carbon structures  

NASA Astrophysics Data System (ADS)

This paper presents a review of results in the area of fission product sorption in carbon structures. Emphasis is placed on identifying those parameters of carbon-based materials that likely play a dominant role in fission product sorption and the extent to which these parameters have been studied. In particular, we discuss published studies of the effects of atomic structure, sp2 to sp3 bonding ratio, coke content, defect structures, irradiation level, and percent of amorphous structures and porosity. Furthermore, the evolution of theories and models for carbon sorption are summarized. A review of the literature available to the authors reveals that the mechanics governing fission product sorptivity remain to be fully understood.

Londono-Hurtado, A.; Szlufarska, I.; Bratton, R.; Morgan, D.

2012-07-01

71

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

72

(Fission product transport experiments (HFR-B1))  

SciTech Connect

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

Myers, B.F.

1989-12-05

73

Fission properties and production mechanisms for the heaviest known elements  

SciTech Connect

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

Hoffman, D.C.

1981-01-01

74

Data Summary Report for Fission Product Release Test HI-1.  

National Technical Information Service (NTIS)

The first in a series of high-temperature fission product release tests was conducted for 30 min at 1400C, with the release taking place into flowing steam. The fuel specimen was a 20-cm-long section of H.B. Robinson fuel rod, irradiated to 28,000 MWd per...

M. F. Osborne R. A. Lorenz J. R. Travis C. S. Webster

1982-01-01

75

Studies of fission product movement in tuffaceous media  

Microsoft Academic Search

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

Thompson

1991-01-01

76

Studies of fission product movement in tuffaceous media.  

National Technical Information Service (NTIS)

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 cons...

J. L. Thompson

1991-01-01

77

Recent MELCOR and VICTORIA Fission Product Research at the NRC  

SciTech Connect

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes in the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.

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

1999-01-21

78

HYPERFUSE: A Hypervelocity Inertial Confinement System for Fusion Energy Production and Fission Waste Transmutation.  

National Technical Information Service (NTIS)

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 ...

H. Makowitz J. R. Powell R. Wiswall

1980-01-01

79

HYPERFUSE: A Hypervelocity Inertial Confinement System for Fusion Energy Production and Fission Waste Transmutation.  

National Technical Information Service (NTIS)

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 I...

H. Makowitz J. R. Powell R. Wiswall

1980-01-01

80

Thermodynamics of the System Uranium Dioxide-Fission Products-Zircaloy.  

National Technical Information Service (NTIS)

Equilibrium thermodynamic calculations were performed in the UO sub 2 -Fission products-Zry (Zircaloy) System under LWR conditions. The oxygen potential gradient of the fuel sets the chemical states of the fission products. Inspection of the standard free...

R. Kohli

1978-01-01

81

Design of an Online, Multispectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor  

Microsoft Academic Search

The U.S. Department of Energy is embarking on a series of tests of tristructural isotropic (TRISO) coated-particle reactor fuel for the Advanced Gas Reactor (AGR). As one part of this fuel development program, a series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory's (INLs) advanced test reactor (ATR). The first test in this series (AGR-1)

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

2007-01-01

82

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

SciTech Connect

The release of fission gas (Xe and Kr) and helium out of nuclear fuel materials in normal operation of a nuclear power reactor can constitute a strong limitation of the fuel lifetime. Moreover, radioactive isotopes of Xe and Kr contribute significantly to the global radiological source term released in the primary coolant circuit in case of accidental situations accompanied by fuel rod loss of integrity. As a consequence, fission gas release investigation is of prime importance for the nuclear fuel cycle economy, and is the driven force of numerous R and D programs. In this domain, for solving current fuel behavior understanding issues, preparing the development of new fuels (e.g. for Gen IV power systems) and for improving the modeling prediction capability, there is a marked need for innovations in the instrumentation field, mainly for: . Quantification of very low fission gas concentrations, released from fuel sample and routed in sweeping lines. Monitoring of quick gas release variations by quantification of elementary release during a short period of time. Detection of a large range of atomic masses (e.g. H{sub 2}, HT, He, CO, CO{sub 2}, Ne, Ar, Kr, Xe), together with a performing separation of isotopes for Xe and Kr elements. Coupling measurement of stable and radioactive gas isotopes, by using in parallel mass spectrometry and gamma spectrometry techniques. To fulfill these challenging needs, a common strategy for analysis equipment implementation has been set up thanks to a recently launched collaboration between the CEA and the Univ. of Provence, with the technological support of the Liverpool Univ.. It aims at developing a chronological series of mass spectrometer devices based upon mass filter and 2D/3D ion traps with Fourier transform operating mode and having increasing levels of performances to match the previous challenges for out-of pile and in-pile experiments. The final objective is to install a high performance online mass spectrometer coupled to a gamma spectrometer in the fission product laboratory of the future Jules Horowitz Material Test Reactor. An intermediate step will consist of testing first equipment on an existing experimental facility in the LECA-STAR Hot Cell Laboratory of the CEA Cadarache. This paper presents the scientific and operational stakes linked to fission gas issues, resumes the current state of art for analyzing them in nuclear facilities, then presents the skills gathered through this collaboration to overcome technological bottlenecks. Finally it describes the implementation strategy in nuclear research facilities of the CEA Cadarache. (authors)

Zerega, Y.; Reynard-Carette, C. [Univ. of Provence, Laboratoire Chimie Provence, UMR 6264, Avenue escadrille Normandie - Niemen, F-13397 Marseille (France); Parrat, D. [CEA, Nuclear Energy Div. DEN, CEA Cadarache, F-13108 Saint-Paul-lez-Durance (France); Carette, M. [Univ. of Provence, Laboratoire Chimie Provence, UMR 6264, Avenue escadrille Normandie - Niemen, F-13397 Marseille (France); Brkic, B. [Univ. of Liverpool, Dept. of Electrical Engineering and Electronics, Liverpool L69 3BX (United Kingdom); Lyoussi, A.; Bignan, G. [CEA, Nuclear Energy Div. DEN, CEA Cadarache, F-13108 Saint-Paul-lez-Durance (France); Janulyte, A.; Andre, J. [Univ. of Provence, Laboratoire Chimie Provence, UMR 6264, Avenue escadrille Normandie - Niemen, F-13397 Marseille (France); Pontillon, Y.; Ducros, G. [CEA, Nuclear Energy Div. DEN, CEA Cadarache, F-13108 Saint-Paul-lez-Durance (France); Taylor, S. [Univ. of Liverpool, Dept. of Electrical Engineering and Electronics, Liverpool L69 3BX (United Kingdom)

2011-07-01

83

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

NASA Astrophysics Data System (ADS)

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

Rest, J.

1985-04-01

84

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

85

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

Microsoft Academic Search

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

S. E. Fisher; K. E. Lenox

1995-01-01

86

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

SciTech Connect

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

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

1987-01-01

87

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

88

Transmutation of Fission Products and Transuranium by High Energy Neutron  

Microsoft Academic Search

We discuss the feasibility of the system that incinerate radioactive fission products (FP) and transuranium (TRU) by using high energy neutrons. As high energy neutron sources, pCF reaction, fusion reaction, and spallation reaction were investigated. In the system that utilizes pCF reaction, a subcritical core made of FP and TRU is bombarded by 14 MeV neutron generated via pCF reaction.

Hideo Haradal; Hiroshi Takahashil Arnold Aronsonl; Kenji Konashi; Takeshi Kase; Nobuyuki Sasao

89

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

90

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.

91

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

92

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

93

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

94

A preliminary comparison of gas core fission and inertial fusion for the space exploration initiative  

Microsoft Academic Search

Potential utilization of fission and fusion-based propulsion systems for solar system exploration is examined using a Mars mission as basis. One system employs the open cycle gas core fission reactor (GCR) as the energy source, while the other uses the fusion energy produced in an inertial Confinement Fusion (MICF) concept, to convert thermal energy into thrust. It is shown that

Terry Kammash; David L. Galbraith

1992-01-01

95

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

SciTech Connect

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

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

2011-07-01

96

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

97

Microstructure and fission gas bubbles in irradiated mixed carbide fuels at 2 to 11 a/o burnup  

NASA Astrophysics Data System (ADS)

An analysis of the defect structure and of small fission gas bubbles has been performed on mixed carbide fuels with burn-ups between 1.8 and 11 a/o by transmission electron microscopy (TEM). A complex defect structure consisting of dislocations, loops and at least 3 types of solid fission product precipitates was observed. Na-bonded carbides develop predominantly a dislocation network increasing in density with burn-up whereas He-bonded carbides showed mainly a corresponding network of crystallographic needle precipitates. Locally the nucleation and growth of small fission gas bubbles with 1 to 20 nm diameters (bubble population P 1) is closely related to their dislocation or needle environment, larger bubbles with diameters 30 to 50 nm appear to be mostly associated with platelike precipitates or dislocation boundaries. The local swelling contribution ? 1 of bubble population P 1 is ? 0.5% and its fission gas content G 1 is 4 to 5% of the total amount of gas created over the whole burn-up range investigated.

Ray, I. L. F.; Blank, H.

1984-05-01

98

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

99

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

100

Evolution of fission-gas-bubble-size distribution in recrystallized U10Mo nuclear fuel  

Microsoft Academic Search

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles, used to characterize fission-gas bubble development in UMo alloy fuel with burnup limited to less than 10at.% U in order to capture the fuel swelling stage prior to irradiation-induced recrystallization, is extended to recrystallized fuel at a burnup of ?16at.% U. During recrystallization the grain size

J. Rest

2010-01-01

101

An initial assessment of a mechanistic model, GRASS-SST, in U-Pu-Zr metallic alloy fuel fission-gas behavior simulations  

NASA Astrophysics Data System (ADS)

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 investigate 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 evaluate fission gas release, fission gas and fission product induced swelling, and detailed gas bubble size distributions in three different fuel zones: the outer ?-U, the intermediate, and the inner ?-U zones. Due to its special microstructural features, the ?-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 in reasonable consistence 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 ?-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 ?-U zone of U-Pu-Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreement 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. Material properties such as fuel grain size and thermal diffusivity of gas and model parameters such as di-atom nucleation probability and gas bubble re-solution constant are predicted by these comparisons. In addition, the results of parametric studies for several parameters are presented for both the intermediate zone and the ?-U zone simulations in order to clarify the sensitivities of simulation results on these parameters.

Yun, Di; Rest, Jeffrey; Hofman, Gerard L.; Yacout, Abdellatif M.

2013-04-01

102

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations  

SciTech Connect

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yusung-gu, Taejon (Korea, Republic of)

2005-05-24

103

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations  

NASA Astrophysics Data System (ADS)

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

2005-05-01

104

A model for fission-gas-bubble behavior in amorphous uranium silicide compounds  

Microsoft Academic Search

A model for the behavior of fission gas in irradiated amorphous materials is developed. The model proposes that gas bubble nucleation occurs within shear bands initiated around free volume regions. Small gasatom clusters that form within these regions are susceptible to dissolution by forces generated by the plastic flow of material around the cluster. The bubble coarsening process depends on

J. Rest

2004-01-01

105

Fission product yield evaluation for the USA evaluated nuclear data files  

SciTech Connect

An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set.

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

1994-10-01

106

Gas production apparatus  

Microsoft Academic Search

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

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

1976-01-01

107

A proposed standard on medical isotope production in fission reactors  

SciTech Connect

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

Schenter, R. E. [Smart Bullets Inc., 2521 SW Luradel Street, Portland, OR 97219 (United States); Brown, G. J. [Ozarks Medical Center, Cancer Treatment Center, Shaw Medical Building, 1111 Kentucky Avenue, West Plains, MO 65775 (United States); Holden, C. S. [Thorenco LLC, 369 Pine Street, San Francisco, CA 94104 (United States)

2006-07-01

108

Fission Product Removal Analysis in APR1400 Containment  

SciTech Connect

In contrast to the TID-14844, NUREG-1465 source term is characterized by realistic and physical behavior of radioactivity in containment. It specifies a categorized release in terms of phenomenological accidental phase and also defines that the dominant form of the fission product iodine is as finely divided airborne particulate. Following a LOCA, radioactive fission products released to containment are primarily in the form of aerosol. In this paper, the aerosol removal mechanisms applied to the APR1400 are discussed. The principal processes that remove aerosols from the containment atmosphere fall into two classes: containment sprays and natural removal processes. Containment sprays are effective in reducing the airborne concentration of particulate iodines as well as other particulates. And also, the aerosol removal rates using the STARNAUA computer program are presented. The results indicate that the NUREG model is very conservative relative to the STCP, a decrease in droplet diameter increases the aerosol removal coefficient, and SRP method underestimates the aerosol removal coefficient in case of smaller aerosol size. (authors)

Jang, Young-Sik; Kim, Tae-Yoon; Ko, Hee-Jin; Lim, Jae-Young; Ko, Kab-Seok [Korea Power Engineering Company, Inc, 360-9 Ma-Buk Ri, Gu-Sung Yup, Yong-In City, Kyung-Ki Do, 449-713 (Korea, Republic of)

2004-07-01

109

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

110

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

Microsoft Academic Search

A new concept for the transmutation of fission products and transuranics is studied. This concept, termed HYPERFUSE, allows one inertial reactor to transmute objectionable fission products (¹³⁷Cs and ⁹°Sr) from a large number (e.g., approximately 30) of light water fission reactors, while at the same time generating electric power from the HYPERFUSE plant at a reasonable net plant efficiency (e.g.,

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

1981-01-01

111

Analysis of fission gas disposition in light water reactor steady-state operation  

SciTech Connect

A model to predict fission gas behavior in irradiated uranium dioxide fuel during the steadystate operation of a nuclear reactor is developed. The basic physical phenomena encountered in analyzing the disposition of fission gas have been retained, but in a simplified form for ease of calculation. The analysis includes treatment of intragranular, grain face, and grain edge gas, and release to open spaces. The code is utilized to obtain comparison with experimental data and to perform fuel behavior studies. The sensitivity studies indicate the importance of grain face and grain edge bubble treatments in modeling fission gas. It is found that representation of release in different sections of the fuel pin is possible in a simple way by assuming evenly spaced bubbles on the edge, and that grain edge bubble interlinkage is a necessary condition for release to the open spaces. The sensitivity studies show that fission gas swelling is mainly due to grain edge bubbles. Grain face bubbles, although large in size, are few in number and contribute little to swelling. Intragranular swelling is intermediate between these two values. The code is successfully used to analyze the Westinghouse fission gas release data from the Zorita, Spain, light water reactor and data from the U.K. reactor DIDO. This success in modeling experiments suggests that the present code can be used in predicting fuel element performance, which is necessary in nuclear fuel design, safety analysis, and interpretation of experimental data on fuel element behavior.

Villalobos, A.; Okrent, D.; Wazzan, A.R.

1982-09-01

112

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

113

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

114

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

115

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

SciTech Connect

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

Skim, Y. S.

1999-11-12

116

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

SciTech Connect

A selective review is presented of previous measurements and the analysis of experiments on the effect of water vapor on fission gas release from uranium oxycarbide. Evidence for the time-dependent composition of the uranium oxycarbide fuel; the diffusional release of fission gas; and the initial, rapid and limited release of stored fission gas is discussed. In regard to the initial, rapid release of fission gas, clear restrictions on mechanistic hypotheses can be deduced from the experimental data. However, more fundamental experiments may be required to establish the mechanism of the rapid release.

Myers, B.F.

1994-04-01

117

Gamma-Ray Spectra of Fission Products Observed with Lithium Drifted Germanium Detectors  

Microsoft Academic Search

The ?-ray spectra of fission products from thermal neutron irradiation of natural U were observed with an encapsulated Ii-drift-Ge ?-ray spectrometer. The spectra were recorded at various periods after irradiation1, 2, 5, 12, 30, 250 days and 3 years. The ?-ray spectra of eight individual fission products including Np were also studied. Due to the high resolution obtained with the

Sei-ichi TAKAYANAGI; Noboru OI; Tetsuji KOBAYASHI; Tohru SUGITA

1966-01-01

118

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

Microsoft Academic Search

The examination of these samples indicated a correlation between the posttest fuel microstructure and the fission product release during the test. As expected, structural changes in the fuel and fission product release increased with test temperature. The effect of steam flow rate, which controls the extent of cladding oxidation, however, was less clear. The amount of fuel-cladding reaction and liquefaction

R. V. Strain; M. F. Osborne

1983-01-01

119

Mass spectrometric measurements of fission product effusion from irradiated light water reactor fuel  

Microsoft Academic Search

Laboratory measurements of fission products effusion from irradiated light water reactor fuels are being carried out at the Joint Research Centre (JRC) of the European Commission in Karlsruhe. The aim of these experiments is twofold: first, data are obtained on diffusion of gaseous and less volatile fission products, which are suitable for a mechanistic analysis of their migration processes in

F. Capone; J. P. Hiernaut; M. Martellenghi; C. Ronchi

1996-01-01

120

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

Microsoft Academic Search

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses

I. C. Gauld; D. E. Mueller

2005-01-01

121

Sensitivity Analysis of Fission Product Concentrations for Light Water Reactor Burned Fuel  

Microsoft Academic Search

The accurate prediction of fission product concentrations (FPCs) is necessary for application of the burnup credit to nuclear facilities. In order to specify important nuclear data for the accurate prediction of FPC, we extensively evaluate the sensitivities of FPC to nuclear data with the depletion perturbation theory. The target fission products are twelve important ones for the burnup credit, Mo-95,

Go CHIBA; Keisuke OKUMURA; Akito OIZUMI; Masaki SAITO

2010-01-01

122

The influence of core degradation phenomena on in-vessel fission product behavior during severe accidents  

Microsoft Academic Search

In-vessel core degradation phenomena influence where fission products will be located and in what chemical forms they will exist and with what materials they will be associated at the time the lower vessel fails in an unmitigated accident sequence. Fission products released from the reactor vessel during the in-vessel phase of core melt progression in a severe reactor accident can

R. R. Hobbins; D. J. Osetek; D. A. Petti; D. L. Hagrman

1988-01-01

123

Steady-State and Transient Fission Gas Release and Swelling Model for LIFE-4.  

National Technical Information Service (NTIS)

The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIF...

A. Villalobos Y. Y. Liu J. Rest

1984-01-01

124

The effect of re-solution models on fission gas disposition in irradiated UO fuel  

Microsoft Academic Search

A computer code developed earlier by Villalobos et al. to predict fission gas behavior in uranium oxide fuel under steady-state irradiation conditions and where bubble gas resolution is represented with the single knock-on model (SKO) is modified to replace the SKO model with the complete bubble destruction model (CBD). The CBD model required that bubble nucleation be included in the

A. R. Wazzan; D. Orkent; A. Villalobos

1985-01-01

125

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

126

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

PubMed

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

Matsson, I; Grapengiesser, B; Andersson, B

2007-01-01

127

Plutonium and surrogate fission products in a composite ceramic waste form.  

SciTech Connect

Argonne National Laboratory is developing a ceramic waste form to immobilize salt containing fission products and transuranic elements. Preliminary results have been presented for ceramic waste forms containing surrogate fission products such as cesium and the lanthanides. In this work results from scanning electron microscopy/energy dispersive spectroscopy and x-ray diffraction are presented in greater detail for ceramic waste forms containing surrogate fission products. Additionally, results for waste forms containing plutonium and surrogate fission products are presented. Most of the surrogate fission products appear to be silicates or aluminosilicates whereas the plutonium is usually found in an oxide form. There is also evidence for the presence of plutonium within the sodalite phase although the chemical speciation of the plutonium is not known.

Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T.

1999-05-19

128

First-principles study of the stability of fission products in uranium monocarbide  

NASA Astrophysics Data System (ADS)

The incorporation and stability of fission products in uranium monocarbide are studied by means of Density Functional Theory using the generalized gradient approximation and projector-augmented waves method. The computations are performed considering incorporation sites of UC, such as the U, C and interstitial sites, and Schottky defects. The computed incorporation energies are discussed on the basis of the atomic size of the fission products, their chemical environment and the electronic structure. These energies show that all the studied fission products would preferentially occupy the U site. However, incorporation energies do not provide any further information on the fission product location in the case of unavailability of the sites which is why the concept of solution energies is also used. The solution energies obtained confirm that all the fission products are expected to be more stable on a U site of a single uranium vacancy or within a non-bound Schottky defect in equilibrium conditions.

Bvillon, mile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

2012-07-01

129

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

Microsoft Academic Search

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

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

2011-01-01

130

HYPERFUSE: A Novel Inertial Confinement System Utilizing Hypervelocity Projectiles for Fusion Energy Production and Fission Waste Transmutation.  

National Technical Information Service (NTIS)

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 ...

H. Makowitz J. R. Powell R. Wiswall

1980-01-01

131

Hyper Fuse: A Novel Inertial Confinement System Utilizing Hypervelocity Projectiles for Fusion Energy Production and Fission Waste Transmutation.  

National Technical Information Service (NTIS)

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 ...

H. Makowitz J. R. Powell R. Wiswall

1979-01-01

132

Modeling of molten core-concrete interactions and fission-product release  

SciTech Connect

The study of molten core-concrete interaction is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program which models the thermal, chemical, and physical phenomena associated with molten core-concrete interactions. Models have been added to extend and improve the modeling of these phenomena. An ideal solution chemical equilibrium methodology is presented to predict the fission-product vaporization release. Additional chemical species have been added, and the calculation of chemical equilibrium has been expanded to the oxidic layer and to the mixed layer configuration. Recent experiments performed at Argonne National Laboratory are compared to CORCON predictions of melt temperature, erosion depth, and release fraction of fission products. The results consistently underpredicted the melt temperatures and erosion rates. However, the predictions of release of Te, Ba, Sr, and U were good. A sensitivity study of the effects of initial temperature, concrete type, use of the mixing option, degree of zirconium oxidation, cavity size, and amount of control material on erosion, gas production, and release of radioactive materials was performed for a PWR and a BWR. The initial melt temperature had the greatest effect on the results of interest. Concrete type and cavity size also had important effects. 78 refs., 35 figs., 40 tabs.

Norkus, J.K.; Corradini, M.L. (Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics)

1991-09-01

133

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

134

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

135

Determining isotopic distributions of fission products with a Penning trap  

NASA Astrophysics Data System (ADS)

A novel method to determine independent yields in particle-induced fission employing the ion guide technique and ion counting after a Penning trap has been developed. The method takes advantage of the fact that a Penning trap can be used as a precision mass filter, which allows an unambiguous identification of the fission fragments. The method was tested with 25MeV and 50MeV proton-induced fission of 238U . The data is internally reproducible with an accuracy of a few per cent. A satisfactory agreement was obtained with older ion guide yield measurements in 25MeV proton-induced fission. The results for Rb and Cs yields in 50MeV proton-induced fission agree with previous measurements performed at an isotope separator equipped with a chemically selective ion source.

Penttil, H.; Karvonen, P.; Eronen, T.; Elomaa, V.-V.; Hager, U.; Hakala, J.; Jokinen, A.; Kankainen, A.; Moore, I. D.; Perjrvi, K.; Rahaman, S.; Rinta-Antila, S.; Rubchenya, V.; Saastamoinen, A.; Sonoda, T.; yst, J.

2010-04-01

136

A model of fission gas behavior during steady-state operation  

SciTech Connect

A model of fission gas behavior during the steady-state operation of a nuclear reactor that uses uranium dioxide as fuel is developed. The basic physical phenomena encountered in analyzing the disposition of fission gas have been retained, but in a simplified form for ease of calculation. The analysis code, includes treatment of intragranular, grain face, and grain edge gas and release to the open spaces. The code is utilized to obtain comparisons with experimental data and to perform fuel behavior sensitivity studies. The results obtained in the sensitivity studies indicate the importance of including grain face and grain edge bubbles treatments in modeling fission gas. It is found that representation of release in different sections of the fuel pin is possible in a simple way by assuming evenly spaced bubbles on the edge, and that grain edge bubble interlinkage is a necessary condition for release to the open spaces. It is also indicated by the sensitivity studies that fission gas swelling is mainly due to grain edge bubbles. Grain face bubbles, although large in size, are few in number and contribute little to swelling. Intragranular swelling is intermediate between these two values. The resulting code can be used in predicting fuel element performance, that is necessary in nuclear fuel design, safety analysis, and interpretation of experimental data on fuel element behavior.

Villalobos, A.

1981-01-01

137

Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation  

SciTech Connect

Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements.

Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

1985-10-01

138

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

139

Determination of iodine-129 in mixed fission products by neutron activation analysis  

SciTech Connect

This report describes an improved method for analyzing /sup 129/I in fission product mixtures originating from fuel reprocessing studies. The method utilizes conventional iodine valence adjustment and solvent extraction techniques to chemically separate /sup 129/I from most fission products. The /sup 129/I is then determined by neutron irradiation and measurement of the 12.4-h /sup 130/I produced by the neutron capture reaction. Special techniques were devised for neutron irradiation of /sup 129/I samples in the pneumatic tube irradiation facilities at the High Flux Isotope (HFIR) and Oak Ridge Research (ORR) reactors. Chemically separated /sup 129/I is adsorbed on an anion exchange resin column made from an irradiation container. The loaded resin is then irradiated in either of the pneumatic facilities to produce /sup 130/I. Sensitivity of the analysis with the HFIR facility (flux: 5 x 10/sup 14/ neutrons cm/sup -2/s/sup -1/) and a 100-s irradiation time is approximately 2 ng. Samples up to 250 mL in volume can be easily processed. The method has been in routine use for about two years and has given good results on samples of reactor fuel solutions and off-gas traps.

Bate, L.C.; Stokely, J.R.

1980-10-01

140

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

141

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

SciTech Connect

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

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

1991-11-01

142

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

143

Baseline Glass Development for Combined Fission Products Waste Streams  

SciTech Connect

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

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

2009-06-29

144

Isomeric yield ratios of fission products in the system of 24 MeV proton-induced fission of 238 U  

Microsoft Academic Search

Isomeric yield ratios of 30 fission products in 24 MeV proton-induced fission of238U were measured by the use of the ion-guide isotope separator on-line. The obtained isomeric yield ratios were converted to the angular momenta of primary fission fragments based on the statistical model. The deduced angular momenta were examined from various aspects. It is found that in general the

M. Tanikawa; H. Kudo; H. Sunaoshi; M. Wada; T. Shinozuka; M. Fujioka

1993-01-01

145

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

Microsoft Academic Search

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

Albert J. Juhasz

2005-01-01

146

Inherent safety phenomenon of fission-gas induced axial extrusion in oxide and metal fueled LMFBRs  

Microsoft Academic Search

The current emphasis in LMFBR design is to develop reactor systems that contain as many features as possible to limit the severity of hypothetical accidents and provide the maximum time before corrective action is required while maintaining low capital costs. One feature is the possibility of fission-gas induced axial extrusion of the fuel within the intact cladding. The potential exists

K. J. Miles; Kalimullah

1985-01-01

147

Overview of Experimental Support for Fission-Product Transport Analyses at Oak Ridge National Laboratory.  

National Technical Information Service (NTIS)

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions...

R. P. Wichner

1983-01-01

148

Fission Product Release Behaviour after Reactor Shutdown: Experience at Tarapur Atomic Power Station.  

National Technical Information Service (NTIS)

A study on the fission product release behaviour after reactor shutdown has been carried out at Tarapur Atomic Power Station. The activity concentrations of iodines, caesium and technetium in reactor water have been measured. The release rates and release...

S. V. Narasimhan G. Venkateswaran K. S. Venkateswarlu

1977-01-01

149

Analysis of the Chemical State of Plutonium and Fission Products in Process Feed Solutions.  

National Technical Information Service (NTIS)

The chemical states of plutonium and fission products in reprocessing feed solutions are discussed and the need for simple routine analytical procedures for the characterization of species with different extractability is stressed. The development of an e...

M. Bonnevie-Svendsen V. Martini

1966-01-01

150

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

SciTech Connect

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

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

2004-07-01

151

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

SciTech Connect

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

Dollins, C.C.

1981-02-01

152

TRIGA fuel enrichment verification based on the measurement of short-lived fission products  

Microsoft Academic Search

A method is developed to verify the 235U content of TRIGA fresh fuel using gamma-ray spectrometry of the short-lived fission products 97Zr\\/97Nb, 132I and 140La. The short-lived fission-product activities can be established by irradiating the fuel in a nuclear reactor. Based on the measured activities, the 235U content can be deduced by iterative calculations. The aim of this work is

Jinn-Jer Peir; Tien-Ko Wang; Chao-Chin Liu

1999-01-01

153

Separation of fission and corrosion products from boric acid solutions by solvent extraction  

Microsoft Academic Search

Extractive purification of boric acid from radioactive corrosion and fission products dissolved in aqueous solutions modelling\\u000a nuclear reactor coolants has been studied. Aliphatic 1,3-diols containing 8 and 9 carbon atoms per molecule were used as extractants\\u000a fro boric acid. The behaviour of some representative corrosion and fission products as well as various factors affecting their\\u000a distribution between the organic and

J. Narbutt; J. Olza; Z. Przyby?owicz; S. Siekierski

1979-01-01

154

Yield of Photo-Neutrons from U235 Fission Products in Heavy Water  

Microsoft Academic Search

The photo-disintegration of the deuteron has been used to study the hard gamma-rays emitted by fission products of U2351. The neutrons created in the process were used as the indicator of the presence of hard gamma-rays. The fission products were placed at the center of a 10'' radius sphere of heavy water. Conclusions about the periods and yields of the

S. Bernstein; W. M. Preston; G. Wolfe; R. E. Slattery

1947-01-01

155

Extraction-chromatographic separation of uranium from long-lived fission products using tributyl phosphate  

Microsoft Academic Search

Extraction-chromatographic separation of uranium from fission products was performed using undiluted tributyl phosphate sorbed\\u000a on Chromosorb W as a stationary phase, and nitric acid (1: 3) as a mobile phase. Most of the fission products that contributed\\u000a greatly to the radiation level of the sample passed through the column; this effected considerable decontamination. Uranium\\u000a retained on the column was quantitatively

N. Tamura; C. Yonezawa

1974-01-01

156

Vapor transport of fission products in postulated severe light water reactor accidents  

SciTech Connect

A methodology based on chemical thermodynamics has been developed to treat the transport of volatile fission products (FPs) through the core and the primary system. The FPs considered are cesium, iodine, tellurium, strontium, and ruthenium, which may pose the major biohazard in postulated severe accidents in light water reactors. The vapor transport of FPs depends on the volatilities of the chemical compounds that are formed in the carrier gas environment in which the FPs are released and transported. Chemically stable forms were evaluated by minimizing the total free energies of the FP/ fuel/gas environment systems. Many gaseous species for each FP were considered and their partial pressures calculated over a range of temperatures (600 to 3000K), the carrier gas environments (total pressure and ratio of H/sub 2//H/sub 2/O), and the total amount of FPs in the system. It was found that the major dependence of the concentration of the FPs was on the gas temperature, and a model was developed to predict the source of volatile FPs. The model showed that the FPs leaving the core region would condense in the cooler regions of the upper plenum and/or the primary system either on the cold surfaces or be transported further as aerosols.

Cubicciotti, D.; Sehgal, B.R.

1984-05-01

157

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

158

Towards a multiscale approach for assessing fission product behaviour in UN  

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

159

Identification and Quantification of Plutonium and Uranium from Fission Product Gamma-Ray Spectra.  

NASA Astrophysics Data System (ADS)

A technique has been developed to distinguish between ^{239}Pu and ^{235}U by observing fission product delayed gamma-rays produced by fissions induced by an external neutron source. If the number of induced fissions per source neutron per unit mass can be determined from Monte Carlo simulation, the material can also be quantified. Trials were performed with yellowcake, HEU-metal, and Pu-metal samples using a TRIGA reactor and a large ^{252}Cf source as neutron sources. Fission product gamma-ray spectra were collected using a high-resolution hpGe detector over time intervals ranging from 60 s to 3000 s following the end of irradiation. By virtue of being greatly overdetermined, the identity of the Special Nuclear Material (SNM) can be unambiguously determined with a high degree of confidence in all cases by applying a set of Figure of Merit functions. Identification can be made without regard to the properties of the matrix provided a sufficient number of fissions can be induced within the sample to permit observation of the fission product gamma-rays. Once identified, the SNM can be quantified with an accuracy determined mainly by the ability to accurately model the fission response of the system using Monte Carlo simulation, within 3.8 percent in this study.

Beddingfield, David Harris

160

Development of a Gas Filled Magnet spectrometer coupled with the Lohengrin spectrometer for fission study  

NASA Astrophysics Data System (ADS)

The accurate knowledge of the fission of actinides is necessary for studies of innovative nuclear reactor concepts. The fission yields have a direct influence on the evaluation of the fuel inventory or the reactor residual power after shutdown. A collaboration between the ILL, LPSC and CEA has developed a measurement program on fission fragment distributions at ILL in order to measure the isotopic and isomeric yields. The method is illustrated using the 233U(n,f)98Y reaction. However, the extracted beam from the Lohengrin spectrometer is not isobaric ions which limits the low yield measurements. Presently, the coupling of the Lohengrin spectrometer with a Gas Filled Magnet (GFM) is studied at the ILL in order to define and validate the enhanced purification of the extracted beam. This work will present the results of the spectrometer characterisation, along with a comparison with a dedicated Monte Carlo simulation especially developed for this purpose.

Kessedjian, G.; Chebboubi, A.; Faust, H.; Kster, U.; Materna, T.; Sage, C.; Serot, O.

2013-03-01

161

Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons  

NASA Astrophysics Data System (ADS)

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(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.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

2010-12-01

162

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

163

Analysis of the Fission Yeast Rad3+ Gene Product.  

National Technical Information Service (NTIS)

The fission yeast Rad3 protein is representative of a class of proteins that play a crucial role in genome maintenance in all eukaryotic cell. In cells lacking Rad3, normal cell cycle arrest and DNA repair are not induced in response to damage. As a resul...

C. R. Chapman T. Enoch

2000-01-01

164

Production of fission and activation product isotopes at Sandia National Laboratories  

SciTech Connect

The mission of the Sandia National Laboratories (SNL) Annular Core Research Reactor (ACRR) and the Hot-Cell Facility has recently changed from support of defense and other programs to support of the U.S. Department of Energy (DOE) isotope production and distribution program (IPDP). SNL`s primary role, in support of IPDP, is ensuring a reliable supply of {sup 99}Mo to the U.S. health care system. SNL will also play a role to complement the isotope production of other DOE reactor facilities such as High-Flux Isotope Reactor at Oak Ridge, Tennessee, the High-Flux Beam Reactor at Brookhaven, New York, and the Advanced Test Reactor in Idaho. The unique characteristics that the SNL facilities offer to the IPDP facility capability are simplicity, multiple irradiation locations, ready irradiation space access, and co-located hot-cell facilities capable of processing a short decay fission product stream. The SNL {sup 99}Mo effort is characterized elsewhere, and this paper is intended to describe the production of additional isotopes that can be produced for medical and other uses that should start soon after the {sup 99}Mo capability has been established. Isotope production in the SNL facilities is through fission or by neutron activation.

Coats, R.L. [Sandia National Lab., Albuquerque, NM (United States)

1997-12-01

165

Production of fission and activation product isotopes at Sandia National Laboratories  

SciTech Connect

The mission of the Sandia National Laboratories (SNL) Annular Core Research Reactor (ACRR) and the Hot Cell Facility (HCF) has recently changed from support of Defense and other programs to support of the Department of Energy (DOE) Isotope Production and Distribution Program (IPDP). SNL`s primary role, in support of IPDP, is ensuring a reliable supply of {sup 99}Mo to the US health care system. SNL will also play a role of complementing the isotope production of other DOE Reactor facilities such as High Flux Isotope Reactor (HFIR) at Oak Ridge, Tennessee; High Flux Beam Reactor (HFBR) at Brookhaven, New York, ad Advanced Test Reactor (ATR) in Idaho. The unique characteristics that the SNL facilities offer to the IPDP facility capability are simplicity, multiple irradiation locations, ready irradiation space access and co-located hot cell facilities capable of processing a short decay fission product stream. The SNL {sup 99}Mo effort is characterized elsewhere and this paper is intended to describe the production of additional isotopes for that can be produced medical and other uses planned to start soon after the {sup 99}Mo capability has been established. Isotope production in the SNL facilities is through fission or by neutron activation.

Coats, R.L.

1997-08-01

166

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

167

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

Microsoft Academic Search

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

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

1984-01-01

168

Isomeric yield ratio of fission product 148 Pr in 235 U( n th , f )  

Microsoft Academic Search

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 momentumJrms can not reproduce either present

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

1984-01-01

169

Local fission gas release and swelling in water reactor fuel during slow power transients  

NASA Astrophysics Data System (ADS)

Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7-4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25-40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/ r0 ~ 0.75. Release tunnels were fully developed at r/ r0 ~ 0.55 with the result that gas release was 60-70% at this position.

Mogensen, M.; Walker, C. T.; Ray, I. L. F.; Coquerelle, M.

1985-04-01

170

Gas production apparatus and method  

Microsoft Academic Search

This invention relates generally to the production of gases, and more particularly to the production of tritium gas in a reliable long operating lifetime systems that employs solid lithium to overcome the heretofore known problems of material compatibility and corrosion, etc., with liquid metals. The solid lithium is irradiated by neutrons inside low activity means containing a positive pressure gas

W. E. Winsche; F. T. Miles; J. R. Powell

2009-01-01

171

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

172

Analysis of intergranular fission-gas bubble-size distributions in irradiated uraniummolybdenum alloy fuel  

Microsoft Academic Search

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched UMo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ?7.8 at.% U in order to

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

2009-01-01

173

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

NASA Astrophysics Data System (ADS)

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

Dolan, David J.

174

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

175

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

176

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

177

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

NASA Astrophysics Data System (ADS)

In separate effect tests at 1000-1200 C Ru oxidation rate and content of Ru in escaping air flow have been studied with special emphasis on effects of other fission product elements on the Ru oxidation and transport. The results showed that in the decreasing temperature section (1100-600 C) most of the RuO 3 and RuO 4 (?95%) decomposed and formed RuO 2 crystals; while the partial pressure of RuO 4 in the escaping air was in the range of 10 -6 bar. The re-evaporation of deposited RuO 2 resulted in about 10 -6 bar partial pressure in the outlet gas as well. Measurements demonstrated the importance of surface quality in the decreasing temperature area on the heterogeneous phase decomposition of ruthenium oxides to RuO 2. On the other hand water or molybdenum oxide vapour in air appears to decrease the surface catalyzed decomposition of RuO x to RuO 2 and increases RuO 4 concentration in the escaping air. High temperature reaction with caesium changed the form of the released ruthenium and caused a time delay in appearance of maximum concentration of ruthenium oxides in the ambient temperature escaping gas, while reaction with barium and rare earth oxides extended Ru escape from the high temperature area.

Vr, N.; Matus, L.; Kunstr, M.; Osn, J.; Hzer, Z.; Pintr, A.

2010-01-01

178

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

179

Investigation of the diffusion of atomic fission products in UC by density functional calculations  

NASA Astrophysics Data System (ADS)

Activation energies of U and C atoms self-diffusion in UC, as well as activation energies of hetero-diffusion of fission products (FPs) are investigated by first-principles calculations. According to a previous study which showed a likely U site occupation was favoured for all the FPs, their diffusion is restricted to the uranium sublattice of UC in the present study. In this framework, long-range displacements are only possible through a concerted mechanism with a surrounding uranium vacancy. Using the apparent formation energies of the uranium vacancy defect calculated in our previous study and the classical approach used in UO2 by Andersson et al., the activation energies of the main fission products in the various stoichiometric domains have been calculated. The results are compared to those obtained with the five frequency model applied to two representative fission products, Xe and Zr. Interestingly, despite strong differences of formalism, both models provided similar activation energies.

Bvillon, mile; Ducher, Roland; Barrachin, Marc; Dubourg, Roland

2013-03-01

180

Trapping and diffusion of fission products in ThO 2 and CeO 2  

NASA Astrophysics Data System (ADS)

The trapping and diffusion of Br, Rb, Cs and Xe in ThO 2 and CeO 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 2 and CeO 2 are also calculated. Alkali metals are relatively more mobile than other fission products, and bromine is the least mobile.

Xiao, H. Y.; Zhang, Y.; Weber, W. J.

2011-07-01

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

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

183

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

SciTech Connect

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

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

2007-10-18

184

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

SciTech Connect

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

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

1988-01-01

185

Pyrene degradation by a Mycobacterium sp.: identification of ring oxidation and ring fission products.  

PubMed Central

The degradation of pyrene, a polycyclic aromatic hydrocarbon containing four aromatic rings, by pure cultures of a Mycobacterium sp. was studied. Over 60% of [14C]pyrene was mineralized to CO2 after 96 h of incubation at 24 degrees C. High-pressure liquid chromatography analyses showed the presence of one major and at least six other metabolites that accounted for 95% of the total organic-extractable 14C-labeled residues. Analyses by UV, infrared, mass, and nuclear magnetic resonance spectrometry and gas chromatography identified both pyrene cis- and trans-4,5-dihydrodiols and pyrenol as initial microbial ring-oxidation products of pyrene. The major metabolite, 4-phenanthroic acid, and 4-hydroxyperinaphthenone and cinnamic and phthalic acids were identified as ring fission products. 18O2 studies showed that the formation of cis- and trans-4,5-dihydrodiols were catalyzed by dioxygenase and monooxygenase enzymes, respectively. This is the first report of the chemical pathway for the microbial catabolism of pyrene.

Heitkamp, M A; Freeman, J P; Miller, D W; Cerniglia, C E

1988-01-01

186

Pyrene degradation by a Mycobacterium sp. : Identification of ring oxidation and ring fission products  

SciTech Connect

The degradation of pyrene, a polycyclic aromatic hydrocarbon containing four aromatic rings, by pure cultures of a Mycobacterium sp. was studied. Over 60% of ({sup 14}C)pyrene was mineralized to CO{sub 2} after 96 h of incubation at 24{degree}C. High-pressure liquid chromatography analyses showed the presence of one major and at least six other metabolites that accounted for 95% of the total organic-extractable {sup 14}C-labeled residues. Analyses by UV, infrared, mass, and nuclear magnetic resonance spectrometry and gas chromatography identified both pyrene cis- and trans-4,5-dihydrodiols and pyrenol as initial microbial ring-oxidation products of pyrene. The major metabolite, 4-phenanthroic acid, and 4-hydroxyperinaphthenone and cinnamic and phthalic acids were identified as ring fission products. {sup 18}O{sub 2} studies showed that the formation of cis- and trans-4,5-dihydrodiols were catalyzed by dioxygenase and monooxygenase enzymes, respectively. This is the first report of the chemical pathway for the microbial catabolism of pyrene.

Heitkamp, M.A.; Freeman, J.P.; Miller, D.W.; Cerniglia, C.E. (Food and Drug Administration, Jefferson, AR (USA))

1988-10-01

187

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

SciTech Connect

A reliable and computationally efficient physically based model is developed to study the phenomena which regulate intragranular fission gas behavior in LMFBR uranium carbide fuel under operational conditions. Fission gas atoms diffuse in the grain matrix and continuously precipitate into immobile clusters-fission gas bubble embryos-, by agglomeration of two gas atoms. Embryos may survive and grow into equal size fission gas bubbles by additional capture of gas atoms. Gas atoms are simultaneously knocked out of embryos and bubbles back to the grain matrix by an irradiation-induced re-solution process. Embryos nucleate homogeneously at a net rate which accounts for their thermal dissociation and irradiation-induced destruction. Net bubble growth is due to the concentration of re-solution and gas atom diffusion. Embryos and bubbles are at equilibrium with the grain matrix surface tension and behave in non-ideal fashion. The fission gas microscopic swelling is due to the volume of gas atoms, embryos and bubbles. The as-fabricated intragranular porosity is simulated by a spherical and concentric pore in the grain. The model computational economy draws from a formulation based on appropriate similarities, between the actual situation and that of gas atom undisturbed diffusion in the assumed geometry. The model predicts that a quasi-steady state is ultimately reached by the gas atom and embryo concentrations, which causes quasi-steady leakage rates to pore and grain boundary and quasi-linear increases with time in both microscopic swelling and bubble population, while keeping bubble size rather constant. During this regime most of the grain matrix retained gas is in bubbles. Further, the microscopic swelling depends on the grain matrix retained gas and bubble size, being the former the dominant factor.

Madrid, A.

1980-01-01

188

A?100 fission products in the ^24Mg + ^173Yb reaction  

NASA Astrophysics Data System (ADS)

It is difficult to study moderate spin excitations in stable and neutron-rich nuclei because of the lack of suitable target-projectile combinations for fusion-evaporation reactions. However, such excitations in neutron-rich nuclei can be observed in prompt spectroscopy of fission fragments. Products from the fission of the compound nucleus ^197Pb formed in the ^24Mg + ^173Yb reaction at 134.5 MeV were studied using the Gammasphere array at LBNL. Based on extensive ?-ray coincidence relations, detailed level schemes for many fission products, including previously known isotopes of Sr, Y, Zr, Nb, Mo, Tc, and Ru, have been deduced. Previously known level schemes were extended to higher excitations and transitions were assigned to isotopes for which no spectroscopic information was previously reported, by establishing coincidences with known transitions of the complementary fragments. Moderate spin excitations for specific A?100 isotopes will be presented.

Fotiades, N.; Cizewski, J. A.; Ding, K. Y.; McNabb, D. P.; Archer, D. E.; Becker, J. A.; Bernstein, L. A.; Hauschild, K.; Younes, W.; Clark, R. M.; Fallon, P.; Lee, I. Y.; Macchiavelli, A. O.; MacLeod, R. W.

1998-04-01

189

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles  

NASA Astrophysics Data System (ADS)

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO 2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions.

Barrachin, M.; Dubourg, R.; Kissane, M. P.; Ozrin, V.

2009-03-01

190

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

SciTech Connect

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

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

2008-04-14

191

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

192

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

Microsoft Academic Search

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

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

1981-01-01

193

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

Microsoft Academic Search

Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy were 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,

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

1980-01-01

194

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

Microsoft Academic Search

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

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

1979-01-01

195

Temperature and dose dependence of fission-gas-bubble swelling in U 3Si 2  

NASA Astrophysics Data System (ADS)

Large fission gas bubbles were observed during metallographic examination of an irradiated U 3Si 2 dispersion fuel plate (U0R040) in the Advanced Test Reactor (ATR). The fuel temperature of this plate was higher than for most of the previous silicide-fuel tests where much smaller bubble growth was observed. The apparent conditions for the large bubble growth are high fission density (6.1 10 21 f/cm 3) and high fuel temperature (life-average 160 C). After analysis of PIE results of U0R040 and previous ANL test plates, a modification to the existing athermal bubble growth model appears to be necessary for high temperature application (above 130 C). A detailed analysis was performed using a model for the irradiation-induced viscosity of binary alloys to explain the effect of the increased fuel temperature. Threshold curves are proposed in terms of fuel temperature and fission density above which formation and interconnection of bubbles larger than 5 ? are possible.

Kim, Yeon Soo; Hofman, G. L.; Rest, J.; Robinson, A. B.

2009-06-01

196

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

SciTech Connect

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

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

1989-09-01

197

Transport of Radioactive Fission Products from Arable Land and Pastures to Domestic Animals and Man.  

National Technical Information Service (NTIS)

The transport of the fission products exp 137 Cs and exp 90 Sr through food chains under Swedish agricultural conditions was investigated and transport coefficients established. The expected contents of exp 137 Cs and exp 90 Sr in food items produced in s...

A. Eriksson

1978-01-01

198

Determination of fission product noble metals by inductively coupled plasma atomic emission spectrometry  

SciTech Connect

Since less than 1% of the naturally occurring world supply of ruthenium (Ru), rhodium (Rh), and palladium (Pd) is available from US sources, an alternative unexploited source is the Ru, Rh, and Pd created as fission products during the burnup of nuclear fuel. The Pacific Northwest Laboratory (PNL), is conducting a research program to develop a cost effective, waste management-compatible process for extracting noble metals from defense, and possible commercial, fission product waste. To extract noble metals from fission products using the gold-ore fire assay method, the fission products containing the noble metals, are mixed with lead oxide, a reducing agent such as charcoal or flour, and glass forming materials. The glass forming materials are added so that after the noble metals have been separated, the radioactive waste will be enclosed in a glass matrix suitable for permanent storage. The mixture is fused and the lead oxide is reduced to the molten metal. During this reducing process, the molten lead separates the noble metals from the waste/glass mixture thus forming a lead ingot containing relatively pure noble metals. 5 references, 2 figures, 2 tables.

Lautensleger, A.W.

1983-10-01

199

RARE-EARTH METAL FISSION PRODUCTS FROM LIQUID U-Bi  

Microsoft Academic Search

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

Wiswall

1960-01-01

200

A METHOD FOR DETERMINING MATERIAL ATTRIBUTES FROM POST DETONATION FISSION PRODUCT MEASUREMENTS OF AN HEU DEVICE  

Microsoft Academic Search

An algorithm was developed that uses measured isotopic ratios from fission product residue following the detonation of a nuclear weapon to compute the original attributes of the nuclear material used in the weapon. While more accurate (and more computationally intensive) methods are being explored by others, the method described here could serve as a preprocessing step to a more detailed

Adrienne M. LaFleur; William S. Charlton

201

Data Sheets of Fission Product Release Experiments for Light Water Reactor Fuel, (2).  

National Technical Information Service (NTIS)

This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO sub 2 pellets, photographs of p...

N. Ishiwatari H. Nagai T. Takeda K. Yamamoto C. Nakazaki

1979-01-01

202

Measurements of Fission Products Released in Primary Cooling System of OWL-1 Loop.  

National Technical Information Service (NTIS)

Fission products (FPs) and exp 239 Np were measured during a series of FP release experiments in a high-temperature, high-pressure in-pile water loop (OWL-1) in the JMTR at JAERI. The main chemical form of FP iodine in the loop water was I exp - . Iodate ...

K. Yamamoto I. Yokouchi I. Hisa S. Okagawa N. Ishiwatari

1983-01-01

203

Data Sheets of Fission Product Release Experiments for Light Water Reactor Fuel, (3).  

National Technical Information Service (NTIS)

This report is the last data sheets in series of fission product (FP) release experiments for light water reactor (LWR) fuel. Experimental results in the eighth run (in part), and those in the ninth run in OWL-1 of JMTR, are recorded; photographs of fuel ...

N. Ishiwatari H. Nagai K. Yamamoto T. Hirota H. Itami

1981-01-01

204

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

Microsoft Academic Search

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

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

1983-01-01

205

Data Sheets of Fission Product Release Experiments for Light Water Reactor Fuel, (1).  

National Technical Information Service (NTIS)

A series of fission product (FP) release experiments is proceeding with Oarai Water Loop 1 (OWL-1) installed to the JMTR. The purpose is to clarify behavior and mechanism of FP release from a defective fuel rod in normal operation light water reactor (LWR...

K. Yamamoto C. Nakazaki H. Itami N. Ishiwatari Y. Togo

1979-01-01

206

FPFP 2: A code for following airborne fission products in generic nuclear plant flow paths  

Microsoft Academic Search

In order to assure that a nuclear power plant control room remains habitable during certain types of postulated accidents, Pacific Northwest Laboratory (PNL) has undertaken a special study for the US Nuclear Regulatory Commission. This purpose of this study is to develop software that can aid in the analyses of control room habitability during accidents in which airborne fission products

P. C. Owcarski; K. W. Burk; J. V. Ramsdell; D. D. Yasuda

1991-01-01

207

Effectiveness of Engineered Safety Feature (ESF) systems in retaining fission products: background information  

Microsoft Academic Search

The Pacific Northwest Laboratory has compiled and reviewed base line data on the effectiveness of Engineered Safety Feature (ESF) systems in the retention of fission products and particulate material resulting from a nuclear reactor accident. This work is part of an NRC project to provide the best estimates of the consequences of severe reactor accidents. The resulting report describes the

J. Mishima; D. E. Blahnik; M. A. Halverson; A. K. Postma; F. R. Zaloudek

1984-01-01

208

Radioactive Fission Product Release from the Failed Fuel Elements at the VK-50 Reactor.  

National Technical Information Service (NTIS)

The dynamics of release of radioactive gaseous fission products (GFP) namely exp 133 Xe, exp 135 Xe, sup(135m)Xe, exp 138 Xe, sup(85m)Kr, exp 87 Kr, exp 88 Kr, exp 89 Kr, exp 131 I, exp 132 I, exp 133 I, exp 134 I and exp 135 I from failed fuel elements d...

A. V. Vasilishchuk V. V. Konyashov E. I. Shkokov Y. V. Chechetkin E. K. Yakshin

1982-01-01

209

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...

210

Development of a quantum molecular dynamic (QMD) model to describe fission and fragment production.  

PubMed

QMD model coupled with Generalized Evaporation model by S. Furihata (GEM2) is applied for a description of fission nuclei production in p+U interactions at 100 MeV, and for a description of p+Pb --> Bi+X reactions at 10-200 MeV. A good reproduction of the data has been reached. PMID:16381698

Polanski, A; Petrochenkov, S; Uzhinsky, V; Baznat, M

2005-01-01

211

Study of fragment yields, fission and neutron production in lead targets induced by intermediate energy protons.  

PubMed

A review of the experiments on neutron production in thin and thick lead targets and those involving fission reactions and nuclear fragment emission from various targets performed with proton beams from the Dubna synchrophasotron is given. Two different experimental methods, TOF and SSNTD, were used in the measurements. A dependence of the results on proton energy and target type is discussed. PMID:16604716

Yurevich, Vladimir

2005-01-01

212

Tables of Rcn-2 Fission-Product Cross Section Evaluation. Vol. 1 (24 Nuclides).  

National Technical Information Service (NTIS)

The first part of the RCN-2 evaluation of neutron cross-sections for fission product nuclides contains data for 24 nuclides, i.e. exp 93 Nb, sup(92,94,95,96,97,98,100)Mo, exp 99 Tc, sup(101,102,104)Ru, exp 103 Rh, sup(102,104,105,106)Pd, sup(107,108,110)P...

H. Gruppelaar

1977-01-01

213

ACRR (Annular Core Research Reactor) fission product release tests: ST1 and ST2  

Microsoft Academic Search

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

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

1988-01-01

214

Analysis of fission products a method for verification of a CTBT during on?site inspections  

Microsoft Academic Search

If under a Comprehensive Test Ban Treaty a suspicious event can be accurately located, it should be possible to sample possible fission products through an on?site inspection. Such sampling, focused on analysis of the relative abundance of a few key isotopes, can be used to determine the time of a nuclear explosion to an uncertainty of a few hours or

Li Bin

1998-01-01

215

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

Microsoft Academic Search

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

P. M. Sforza; R. J. Cresci

1997-01-01

216

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

217

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

SciTech Connect

The US and the United Kingdom have been engaged in a joint research program in which samples of higher actinides were irradiated in the 600-MW Dounreay Prototype Fast Reactor in Scotland. Analytical results using mass spectrometry and radiometry for actinides and fission products are now available for the samples in Fuel Pins 1 and 2 which were irradiated for 63 full-power days and for the samples in Fuel Pin 4 which were irradiated for 492 full-power days. Results from these three fuel pins are providing estimates of integral cross sections and fission yields.

Raman, S.; Murphy, B.D.

1995-09-01

218

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

Microsoft Academic Search

The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density graphite-moderated core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory's (INL's) Advanced Test Reactor (ATR). One important measure of fuel performance in

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

2006-01-01

219

UO Sub 2 and Fission Product Release from Sodium Pools.  

National Technical Information Service (NTIS)

In the KfK-NALA-program, the release of uranium, cesium, iodine and strontium from hot sodium into an inert gas atmosphere is investigated. The experiments are related to the SNR 300 core catcher problem. Values for the sodium release rate are important b...

W. Schuetz

1980-01-01

220

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

221

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

NASA Astrophysics Data System (ADS)

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

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

1998-08-01

222

In-core thermal-hydraulic and fission product calculations for severe fuel damage analyses  

SciTech Connect

Best-estimate calculations of realistic fission product source terms are presented for the Severe Fuel Damage (SFD) tests conducted in the Power Burst Facility (PBF), utilizing the Advanced Reactor Severe Accident Program (ARSAP) bulk mass transfer correlation. Computer codes were written to perform the thermal-hydraulic and fission product calculations for the SFD tests. Fewer and slower releases are predicted with the ARSAP mass transfer correlation, in good agreement with the test results. The ARSAP mass transfer model correlates the inverse fuel temperature with the product of release rate and grain size considering the fuel/cladding interaction. The empirical coefficients were developed from Oak Ridge National Laboratory (ORNL) high-burnup fuel data in the 770 to 2,275 K temperature range. The ORNL test data indicate that the fuel/cladding interaction takes effect above 2,000 K.

Suh, K.Y.; Sharon, A.; Hammersley, R.J. (Fauske Associates, Inc., Burr Ridge, IL (USA))

1989-11-01

223

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

224

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

225

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

226

EIA's Natural Gas Production Data  

EIA Publications

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

Information Center

2009-04-09

227

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

228

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

SciTech Connect

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

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

1980-03-01

229

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

SciTech Connect

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

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

1982-07-01

230

Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents  

SciTech Connect

This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

Ellison, P.G.; Monson, P.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Mitchell, H.A. [Concord Associates, Inc., Knoxville, TN (United States)

1990-12-31

231

Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents  

SciTech Connect

This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

Ellison, P.G.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); Mitchell, H.A. (Concord Associates, Inc., Knoxville, TN (United States))

1990-01-01

232

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

Microsoft Academic Search

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

W. Schenk; H. Nabielek

1991-01-01

233

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

234

FPFP 2: A code for following airborne fission products in generic nuclear plant flow paths  

SciTech Connect

In order to assure that a nuclear power plant control room remains habitable during certain types of postulated accidents, Pacific Northwest Laboratory (PNL) has undertaken a special study for the US Nuclear Regulatory Commission. This purpose of this study is to develop software that can aid in the analyses of control room habitability during accidents in which airborne fission products could challenge internal air pathways to the control room. PNL has completed an initial version (FPFP) and final version (FPFP 2) of a software package that can estimate the unsteady-state invasion of quantities of fission products into the control room or any other destination within the nuclear plant via generic internal flow paths. This report consists of three parts: Section 2.0, Technical Bases, describes the flow path components and mechanisms of natural fission product deposition; Section 3.0, FPFP 2 Code Description, describes code organization and the functions of the subroutines; and Section 4.0, Code Operation, discusses details of input requirements, code output, and a sample case demonstration. The appendices consist of an FPFP 2 Fortran code listing, a listing of a code for building input files, forms for building input files, and the sample case input and output files. 7 refs., 3 figs.

Owcarski, P.C.; Burk, K.W.; Ramsdell, J.V.; Yasuda, D.D. (Pacific Northwest Lab., Richland, WA (USA))

1991-03-01

235

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

236

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

237

Effects of burnup on fission product release and implications for severe fuel damage events  

SciTech Connect

Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800/sup 0/K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400/sup 0/K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products.

Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

1984-01-01

238

Novel Fission-Product Separation Based on Room-Temperature Ionic Liquids. (Report for September 15, 2001-September 14, 2004).  

National Technical Information Service (NTIS)

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 seco...

2004-01-01

239

Fuel behaviour and fission product release under realistic hydrogen conditions comparisons between HEVA 06 test results and Vulcain computations.  

National Technical Information Service (NTIS)

HEVA 06 test was designed to simulate the conditions of fission products release from irradiated fuel under hydrogen conditions occurring in a PWR core at low pressure. The test conditions were defined from results provided by the core degradation module ...

J. M. Dumas G. Lhiaubet G. Le Marois G. Ducros

1989-01-01

240

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

NASA Astrophysics Data System (ADS)

The mass-yield distribution of fission products in the 2.5 GeV bremsstrahlung-induced fission of 209Bi have been determined by using the recoil catcher and the off-line ?-spectrometry technique in the high energy electron linac at the Pohang Accelerator Laboratory. The mass-yield determination involves the measurements of cumulative yields for 32 fission products and independent yields of 17 fission products in the photo-fission of 209Bi nuclei. It was found that the mass-yield distribution of fission products in 209Bi is symmetric with an average mass of 95 0.5 and a FWHM of 51 2.0 mass units. Present data at 2.5 GeV along with the literature data at 1 GeV, 700-600 MeV, and 85-28 MeV were interpreted from the point of increase of multi-chance fission and multi-nucleon emission probabilities with an increase in excitation energy. It was found that the average mass of the mass-yield distribution of the fission products decreases from 103 0.5 at 28-85 MeV to 95 0.5 at 2.5 GeV. On the other hand, the FWHM of the mass-yield distribution increases from 19 mass units at 28-40 MeV to 51 mass units at 2.5 GeV. It was also found that the nuclear structure effect observed at the photo-fission of 209Bi with 28-85 MeV bremsstrahlung is washed out at higher energy.

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

2009-06-01

241

On the role of grain boundary diffusion in fission gas release  

NASA Astrophysics Data System (ADS)

It is generally believed that thermal fission gas release from LWR fuel occurs mainly via interconnected grain boundary bubbles. Grain boundary diffusion is not considered to be a significant mechanism. We investigated this supposition by two methods; first, by assessing the distance a gas atom can migrate in a grain boundary containing perfectly absorbing traps. For areal number densities and fractional coverages by the traps observed in fuel irradiated to burnups exceeding 20 MWd/kg, gas atoms will be trapped after a migration distance equal to the size of a grain or less. This supports the supposition for medium-to-high burnups. However, the above-mentioned model is inapplicable for trace-irradiated specimens. In our second analysis, we examined Xe release from trace-irradiated UO 2. The measurements indicated that the liberation involves more than only lattice diffusion at the specimen surface, and that the data are consistent with sequential lattice and grain boundary diffusion unimpeded by intergranular traps. The analysis also provided rough estimates of the grain boundary diffusion coefficient in UO 2.

Olander, D. R.; Van Uffelen, P.

2001-02-01

242

Cation exchange selectivity of some fission products on strongly acidic cation exchanger of sulfonic acid type-nitric acid system  

Microsoft Academic Search

Distribution coefficients of fission products in nitric acid for strongly acidic cation exchanger of sulfonic acid type with different cross-linking and structure were measured by a column method. Uptake of cationic fission products increases with resin cross-linking and decreases of nitric acid concentration. The distribution coefficient of the ion, [KdMn+]*, in a given system is expressed as log [KdMn+]*=Blog[KdMn+]+A where

T. Sato

1990-01-01

243

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

Microsoft Academic Search

Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO2 and UO2 oxides, and the MgO\\/(U,Hf,Ce)O2 interfaces have been carried out. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO2. However, this trend is reversed or reduced for alkaline earth oxides with

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

2010-01-01

244

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

Microsoft Academic Search

Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO2 and UO2 oxides, and the MgO\\/(U, Hf, Ce)O2 interfaces have been carried out. These calculations demonstrate that the fission product incorporation energies in MgO are higher than in HfO2. However, this trend is reversed or reduced for alkaline earth

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

2010-01-01

245

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

PubMed

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

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

2014-03-28

246

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

247

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 accent="true">?e spectrum. The final accent="true">?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

248

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

249

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

250

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

NASA Astrophysics Data System (ADS)

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

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

1996-10-01

251

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

SciTech Connect

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

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

2012-05-30

252

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

253

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

254

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

255

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

256

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

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

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

SciTech Connect

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

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

1984-01-01

259

Production of pipeline gas from coal  

Microsoft Academic Search

A process is disclosed for producing pressurized pipeline gas wherein coal is gasified in oxygen at a relatively low pressure, typically less than 5 atmospheres, to produce a raw gas containing carbon monoxide, hydrogen, carbon dioxide, gaseous sulfur compounds and particulates. A major portion of the raw gas is cooled, cleaned and methanated to produce a pipeline quality product gas

Blaskowski

1983-01-01

260

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

261

Experimental evaluation of fission-gas release in LMFBR subassemblies using an electrically heated test section with sodium as coolant  

Microsoft Academic Search

A description is given of an out-of-pile experiment which simulated ; fission-gas release in current-design uranium-oxide fuel subassemblies of liquid-; metal-cooled fast breeder reactors (LMFBR's) and which was performed to evaluate ; the potential for pin-to-pin failure propagation due to thermal transients ; induced in adjacent fuel pins. A sodium-cooled test section containing three ; electrically heated pins was used.

R. E. Wilson; J. B. van Erp; T. C. Chawla; E. L. Kimont; R. D. Baldwin

1973-01-01

262

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

NASA Astrophysics Data System (ADS)

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

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

1997-09-01

263

Use of fission gas release characteristics to identify type of fuel element failure in LMRs  

SciTech Connect

The performance of liquid-metal reactor (LMR) metal fuel elements has been studied at the Experimental Breeder Reactor II (EBR-II) for >25 yr and valuable fission gas (FG) data have been accumulated from the very occasional fuel element failure. In addition, the run-beyond-clad-breach (RBCB) performance of metal fuel has been specifically studied since 1986. From May 1986 to the present, archived data were obtained from the data acquisition system at EBR-II with the specific intention of studying the FG release behavior of different types of metal fuel breaches. The slope of natural logarithm of release to birth ratio versus the logarithm of decay constant of seven FG isotopes (i.e., {sup 85m}Kr, {sup 87}Kr, {sup 88}Kr, {sup 133}Xe, {sup 135}Xe, {sup 135m}Xe, {sup 138}Xe) as a function of time along with their activity were calculated and plotted and distinct characteristics were observed for different types of failure, e.g., upper plenum versus fuel column failures. A description of the analytical methods used and a comparison of the FG release behavior of these type types of breach are presented in this paper.

Mikaili, R.; Beck, W.N.; Lambert, J.D.B. (Argonne National Lab., IL (United States))

1991-01-01

264

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

265

Monte Carlo models for the production of ?-delayed gamma-rays following fission of special nuclear materials  

NASA Astrophysics Data System (ADS)

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

Pruet, J.; Hall, J.; Descalle, M.-A.; Prussin, S.

2004-08-01

266

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

267

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

SciTech Connect

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

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

1996-10-01

268

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

SciTech Connect

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

Martin, R.C.

1990-01-01

269

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

270

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

SciTech Connect

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

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

1986-05-01

271

Simulation of neutron rich nuclei production through (sup 239)U fission at intermediates energies.  

National Technical Information Service (NTIS)

The theoretical part and some results obtained from a model realised for fission processes in wide range of mass-asymmetries are presented. The fission barriers are computed in a tridimensional configuration space using the Yukawa - plus - exponential mac...

M. Mirea F. Clapier N. Pauwels J. Proust

1997-01-01

272

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

SciTech Connect

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

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

2008-07-15

273

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

274

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

275

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

276

Actinide, Elemental, and Fission Product Measurements by ICPMS at the Savannah River Site  

SciTech Connect

VG Elemental Inductively coupled plasma-mass spectrometer (ICPMS), PlasmaQuad 1 (PQ1) Model No. 4, installed in a radiohood, is used by the Savannah River Technology Center to provide non-routine mass measurements for environmental monitoring, waste tank characterization studies, isotope ratios for criticality determinations, and the measurement of elemental, fission product, and actinide mass distributions of the glass product from the Defense Waste Processing Facility (DWPF). Modifications to improve instrument reliability, sample preparation, and data handling, as well as modifications to the laboratory that permit measurements in a radioactive environment will be discussed. Based on our operating experience, two laboratory facilities are being prepared for additional instruments to operate in a radioactive environment. A separate instrument is being installed for non-radioactive measurements and method development.

Tovo, L.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Waller, P.R.; Clymire, J.; Jones, V.D.; Boyce, W.T.

1998-03-01

277

Rates of defect production by fission neutrons in metals at 4.7 K  

NASA Astrophysics Data System (ADS)

As part of an interlaboratory program, we have measured the resistivity-damage rates at 4.7 K for the dilute alloys, V-300 ppm Zr, Nb-300 ppm Zr, and Mo-300 ppm Zr, irradiated by virtually unmoderated fission neutrons. In addition, Al, Ni, Cu, and stainless steel have also been measured to provide a broader data base for comparison with other experimental work using a variety of neutron spectra and with defect-production theory. A broad view of the results shows that the ratio of experimentally to theoretically determined production rates (damage efficiencies) for various fast-neutron spectra ranges from about 0.25 to 0.50. On the other hand, for a given element various neutron-energy spectra peaked from 1 to 15 MeV give variations in damage-efficiency values of only 7-30%.

Coltman, R. R.; Klabunde, C. E.; Williams, J. M.

1981-09-01

278

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

PubMed

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

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

2012-12-01

279

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

280

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

281

Natural Gas Production and Consumption: 1976.  

National Technical Information Service (NTIS)

Data are presented in tabular form on the salient statistics of natural gas in the United States (1972 to 1976); gross withdrawals and disposition of natural gas in the United States (1975 to 1976); quantity and value of marketed production of natural gas...

1978-01-01

282

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

NASA Astrophysics Data System (ADS)

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

Cubicciotti, Daniel

1985-02-01

283

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

284

Methane Hydrate Gas Production by Thermal Stimulation.  

National Technical Information Service (NTIS)

Two models have been developed to bracket the expected gas production from a methane hydrate reservoir. The frontal-sweep model represents the upper bound on the gas production, and the fracture-flow model represents the lower bound. Parametric studies we...

P. L. McGuire

1981-01-01

285

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

NASA Astrophysics Data System (ADS)

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

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

2009-04-01

286

Effect of fission products on air-oxidation of LWR spent fuel  

NASA Astrophysics Data System (ADS)

Hyperstoichiometric U 4O 9 produced by air-oxidation of LWR spent fuel was studied by ceramography, XRD, TEM, and thermal analysis methods. Several years' oxidation of the spent fuel at 175-195C completely converted as-irradiated pellet fragments and coarse powders to disordered ?-U 4O 9 of composition ~ UO 2.4 without forming U 3O 8. Heating the oxidized fuel to much higher temperatures converted the U 4O 9 to U 3O 8 without forming intermediate oxides such as U 3O 7. Oxidation of impurity-doped UO 2 was also investigated to determine if the fission products in solid solution with the UO 2 in spent fuel could be responsible for the different oxidation behavior of unirradiated and spent fuel. Additions of 4 to 8 wt% Gd 2O 3 in unirradiated UO 2 were also found to stabilize U 4O 9 and delay U 3O 8 formation.

Thomas, L. E.; Einziger, R. E.; Buchanan, H. C.

287

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

288

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

SciTech Connect

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

Ahmad, I.; Morss, L.R. [Argonne National Lab., IL (United States); Durell, J.L. [Manchester Univ. (United Kingdom). Dept. of Physics and Astronomy] [and others

1996-10-01

289

Accident management to prevent containment failure and reduce fission product release  

SciTech Connect

Brookhaven National Laboratory, under the auspices of the US Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The strategies considered make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident, i.e. after vessel breach, and sizable uncertainties exist regarding some of the phenomena involved. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to specific containment types are presented. 2 refs., 5 figs., 2 tabs.

Lehner, J.R.; Lin, C.C.; Luckas, W.J.; Pratt, W.T.

1991-01-01

290

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

NASA Astrophysics Data System (ADS)

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

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

1996-06-01

291

Observation and Measurement of Se-79 in SRS High-Level Tank Fission Product Waste  

SciTech Connect

The authors report the first observation of confirmed Se-79 activity in Savannah River Site high level fission product waste. Se-79 was measured after a seven step chemical treatment to remove interfering activity from Cs-137, Sr-90, and plutonium at levels 105 times higher than the observed Se-79 content and to remove Tc-99 at levels 300 times higher than observed Se-79. Se-79 was measured by liquid scintillation beta-decay counting after specific tests to eliminate uncertainties from possible contributions from Tc-99, Pm-147, Sm-151, Zr-93, or Pu-241, whose beta-decay spectra could appear similar to that of Se-79, and whose content would be expected at levels near or greater than Se-79.

Dewberry, R.A.

2000-08-21

292

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

293

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

PubMed

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

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

2011-08-17

294

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

295

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

296

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

Microsoft Academic Search

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.

Gauntt; Randall O

2010-01-01

297

Preliminary Calculation for Fission Products Generation and Accumulation in Different Types of Fuel Rods by Computer Code FPRM-1.  

National Technical Information Service (NTIS)

The computer code ''FPRM-1'' has been developed for calculation of the quantities of fission products gases released from pellets into plenum in a fuel rod. On the assumption that the irradiation tests of plutonium fuel and others under development in an ...

N. Ishiwatari

1978-01-01

298

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

National Technical Information Service (NTIS)

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code an...

N. A. Hanan

1998-01-01

299

Release Behavior of Fission Products from Coated Fuel Particles During Post-Irradiation Heating at Abnormally High Temperatures.  

National Technical Information Service (NTIS)

The present report describes experimental results on release behavior of metal fission-products during isochronal and isothermal heating of TRISO-and BISO-coated fuel particles at temperatures from 1400 to 2200degC. The particles were irradiated in the ei...

K. Hayashi K. Fukuda

1989-01-01

300

Automated System for Selective Fission Product Separations Applied to the Study of exp 113-115 Pd.  

National Technical Information Service (NTIS)

A microcomputer-controlled radiochemical separation system has been developed for the isolation and study of fission products with half-lives of greater than or equal to 10 s. The system, based upon solvent extraction with three centrifugal contactors cou...

D. H. Meikrantz

1980-01-01

301

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

Microsoft Academic Search

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

John M Scaglione; Don Mueller; John C Wagner

2011-01-01

302

The r Process in the region of transuranium elements and the contribution of fission products to the nucleosynthesis of nuclei with A ? 130  

Microsoft Academic Search

We discuss the influence of nuclear masses and mass distributions of fission products on the formation of heavy elements at\\u000a the final stages of the r-process recycled through fission on long duration timescales. The fission recycling is of great importance in an environment\\u000a with a high density of free neutrons (e.g., in neutron star merger scenarios), when the r-process duration

I. V. Panov; I. Yu. Korneev; F.-K. Thielemann

2008-01-01

303

How EIA Estimates Natural Gas Production  

EIA Publications

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

Information Center

2004-02-01

304

The effect of reusing laser gas on the fission-fragment pumped 1.73 {mu}m atomic xenon laser output  

SciTech Connect

Fission-fragment pumped 1.73 {mu}m atomic xenon laser output was measured without changing the laser gas mixture before each reactor pulse. For a Ar/Xe gas mixture at 260 Torr and 0.3 percent xenon, no degradation in laser output was noted for five reactor pulses.

Hebner, G.A.

1994-08-01

305

Spallation and fission products in the (p+ 179Hf) and (p+ natHf) reactions  

NASA Astrophysics Data System (ADS)

Production of Hf and Lu high-spin isomers has been experimentally studied in spallation reactions induced by intermediate energy protons. Targets of enriched 179Hf (91%) and natHf were bombarded with protons of energy in the range from 90 to 650 MeV provided by the internal beam of the Dubna Phasotron synchrocyclotron. The activation yields of the reaction products were measured by using the ?-ray spectroscopy and radiochemistry methods. The production cross-sections obtained for the 179m2Hf, 178m2Hf and 177mLu isomers are similar to the previously measured values from the spallation of Ta, Re and W targets. Therefore, the reactions involving emission of only a few nucleons, like (p,p'), (p,p'n) and (p,2pn), can transfer high enough angular momentum to the final residual nuclei with reasonable large cross-sections. A significant gain in the isomeric yields was obtained when enriched 179Hf targets were used. The mass distribution of the residual nuclei was measured over a wide range of masses and the fission-to-spallation ratio could be deduced as a function of the projectile energy. Features of the reaction mechanism are briefly discussed.

Karamian, S. A.; Ur, C. A.; Adam, J.; Kalinnikov, V. G.; Lebedev, N. A.; Vostokin, G. K.; Collins, C. B.; Popescu, I. I.

2009-03-01

306

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

NASA Astrophysics Data System (ADS)

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

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

2012-05-01

307

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

308

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

Microsoft Academic Search

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

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

2011-01-01

309

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

310

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

311

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

SciTech Connect

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

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

1991-11-01

312

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

313

The effect of re-solution models on fission gas disposition in irradiated UO/sub 2/ fuel  

SciTech Connect

A computer code developed earlier by Villalobos et al. to predict fission gas behavior in uranium oxide fuel under steady-state irradiation conditions and where bubble gas resolution is represented with the single knock-on model (SKO) is modified to replace the SKO model with the complete bubble destruction model (CBD). The CBD model required that bubble nucleation be included in the analysis. The revised code is used to compute gas release and total swelling. Both are found to be insensitive to whether they are obtained with the CBD or the SKO option. This is mainly because at low atomic percent of burnup, total swelling is dominated by the grain-edge bubble gas contribution, and release is dependent on the formation of a complete grainface/grain-edge tunnel network - factors that are not much affected by either the SKO or CBD models. At higher atomic percent of burnup, intragranular swelling, which can be sensitive to the re-solution model, contributes more to swelling. But even then, computations at 1.0 at .% burnup suggest total swelling will continue to be dominated by grain-edge gas. These results suggest that in modeling swelling and release in irradiated uranium dioxide fuel, the simpler SKO resolution model is satisfactory.

Wazzan, A.R.; Orkent, D.; Villalobos, A.

1985-08-01

314

Development of zirconium/magnesium phosphate composites for immobilization of fission products  

SciTech Connect

Novel chemically bonded phosphate ceramics have been investigated for the capture and stabilization of volatile fission-product radionuclides. The authors have used low-temperature processing to fabricate zirconium phosphate and zirconium/magnesium phosphate composites. A zirconium/magnesium phosphate composite has been developed and shown to stabilize ash waste that has been contaminated with a radioactive surrogate of the {sup 137}Cs and {sup 90}Sr species. Excellent retention of cesium in the phosphate matrix system was observed in both short- and long-term leaching tests. The retention factor determined by the USEPA Toxicity Characteristic Leaching Procedure was one order of magnitude better for cesium that for strontium. The effective diffusivity, at room temperature, for cesium and strontium in the waste forms was estimated to be as low as 2.4 {times} 10{sup {minus}13} and 1.2 {times} 10{sup {minus}11} m{sup 2}/s, respectively. This behavior was attributed to the capture of cesium in the layered zirconium phosphate structure via an intercalation ion-exchange reaction, followed by microencapsulation. However, strontium is believed to be precipitated out in its phosphate form and subsequently microencapsulated in the phosphate ceramic. The performance of these final waste forms, as indicated by the compression strength and the durability in aqueous environments, satisfies the regulatory criteria.

Singh, D. Tlustochowicz, M.; Wagh, A.S. [Argonne National Lab., IL (United States). Energy Technology Div.

1999-01-01

315

A scoping study of fission product transport from failed fuel during N Reactor postulated accidents  

SciTech Connect

This report presents a scoping study of cesium, iodine, and tellurium behavior during a cold leg manifold break in the N Reactor. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The LACE LA1 test and evidence from the Power Burst Facility Severe Fuel Damage tests are used to test the key model applied to determine aerosol behavior. Recommendations for future analysis are also provided. The primary result is that most of the cesium, iodine, and tellurium remains in the molten uranium fuel. Only 0.0035 of the total inventory is calculated to be released. Condensation of the most of the species of cesium and iodine that are released is calculated, with 0.998 of the released cesium and iodine condensing in the spacers and upstream end of the connector tubes. Most of the tellurium that is released condenses, but the chemical reaction of tellurium vapor with surfaces is also a major factor in the behavior of this element.

Hagrman, D.L.

1988-01-01

316

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

317

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

SciTech Connect

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

Takahashi, Hiroshi; Chen, Xinyi

1996-12-31

318

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

SciTech Connect

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

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

1993-10-01

319

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

320

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

PubMed

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

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

2012-10-01

321

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

322

Fission-product gamma-ray line pairs sensitive to fissile material and neutron energy  

NASA Astrophysics Data System (ADS)

The beta-delayed gamma-ray spectra from the fission of 235U, 238U, and 239Pu by thermal and near-14-MeV neutrons have been measured for delay times ranging from 1 min to 14 h. 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.

2008-07-01

323

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

Microsoft Academic Search

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

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

2011-01-01

324

Review Article: The Effects of Radiation Chemistry on Solvent Extraction: 2. A Review of Fission?Product Extraction  

Microsoft Academic Search

The partitioning of the long?lived ??emitters and the high?yield fission products from dissolved nuclear fuel is a key component of processes envisioned for the safe recycling of nuclear fuel and the disposition of high?level waste. These future processes will likely be based on aqueous solvent?extraction technologies for light?water reactor fuel and consist of four main components for the separation of

Bruce J. Mincher; Giuseppe Modolo; Stephen P. Mezyk

2009-01-01

325

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

Microsoft Academic Search

Density functional theory (DFT) calculations of fission product (Xe, Sr, and Cs) incorporation and segregation in alkaline earth metal oxides, HfO and UO oxides, and the MgO\\/(U, Hf, Ce)O interfaces have been carried out. In the case of UO, the calculations were performed using spin polarization and with a Hubbard U term characterizing the on-sit Coulomb repulsion between the localized

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

2009-01-01

326

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

327

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

SciTech Connect

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

Takahashi, H.; Yu, A.

1997-04-01

328

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

329

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

SciTech Connect

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

Brown, D.F.

1994-06-30

330

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

SciTech Connect

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

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

1990-11-01

331

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 418 nuclides; (2) Covariance uncertainty data for 185 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 at higher energies for isotopes of F, 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 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. 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 [1].

G. Palmiotti

2011-12-01

332

Antrim gas play, production expanding in Michigan  

SciTech Connect

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

Not Available

1994-05-30

333

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2010 CFR

...173.350 Combustion product gas. The food additive combustion product gas may be safely used in the...purpose of removing and displacing oxygen in accordance with the following...butane, propane, or natural gas. The combustion...

2010-01-01

334

Fission product release and fuel behavior of irradiated light water reactor fuel under severe accident conditions. The ACRR ST1 Experiment  

Microsoft Academic Search

The annular Core Research Reactor (ACRR) Source Term (ST) Experiment program was designed to obtain time-resolved data on the release of fission products from irradiated fuels under well-controlled light water reactor severe accident conditions. The ST-1 Experiment was the first of two experiments designed to investigate fission product release. ST-1 was conducted in a highly reducing environment at a system

M. D. Allen; H. W. Stockman; K. O. Reil; J. W. Fisk

1991-01-01

335

Three-dimensional simulation of threshold porosity for fission gas release in the rim region of LWR UO 2 fuel  

NASA Astrophysics Data System (ADS)

The threshold porosity above which fission gas release channels would be formed extensively in the rim region of high burnup UO 2 fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cubes, the pore distribution in the rim was simulated three-dimensionally by the Monte Carlo method according to rim porosity and pore size distribution. Using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of the rim porosity. The simulation revealed that it is the rim porosity rather than the pore size distribution or rim thickness that determines the fraction of open pores connected to the pellet surface. The analysis also indicated that the threshold porosity is around 24%, above which the number of rim pores forming release channels increases very rapidly.

Koo, Yang-Hyun; Oh, Je-Yong; Lee, Byung-Ho; Sohn, Dong-Seong

2003-09-01

336

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

SciTech Connect

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

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

1995-03-01

337

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

SciTech Connect

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

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

2000-05-11

338

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

SciTech Connect

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

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

2010-01-01

339

Improving production from tight, geopressured gas zones  

SciTech Connect

Completion experience gained in southwest Texas' Lower Wilcox may have application to other deep, low-permeability, geopressured gas objectives. Core-sample permeabilities have been increased 40-fold, and carefully planned frac treatments have quadrupled production from 17 wells. Although high bottomhole temperatures, flow rates that can exceed 8 MMcfd, and mildly sour production would indicate serious corrosion problems, batch treatment, supplemented by nitrogen displacement, has controlled this potential problem effectively.

Weeks, S.G.

1984-05-01

340

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

Microsoft Academic Search

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

Matthew R. Walsh; Steve H. Hancock; Scott J. Wilson; Shirish L. Patil; George J. Moridis; Ray Boswell; Timothy S. Collett; Carolyn A. Koh; E. Dendy Sloan

2009-01-01

341

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

342

Metal powder production by gas atomization  

NASA Technical Reports Server (NTRS)

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

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

1986-01-01

343

Method for the production of synthesis gas  

SciTech Connect

A method is claimed for the continuous production of synthesis gas comprising of carbon monoxide and hydrogen through the autothermal gasification of solid combustibles in a pressure reactor. The method involves the following: introducing into a screw machine containing two parallely ordered shafts, a finely divided solid combustible; moistening and intimately mixing the solid combustible with 2 to 30% by weight of water, degasing and compressing the moist solid combustible to a pressure higher than that of the reactor; adding the gas-tight compressed and moist solid combustible to a reaction chamber-through a burner where the combustible is brought into contact with the gasification medium; evaporating the water in the compressed and moist solid combustible and producing a comminuted dispersion of the solid combustible in the mixture of the gasification medium and water vapor; reacting the combustible dispersion to give a raw synthesis gas; and removing the raw synthesis gas from the reactor.

Escher, G.; Harjung, J.; Wenning, H.P.

1981-11-24

344

New Methodology for Natural Gas Production Estimates  

EIA Publications

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

Information Center

2010-04-26

345

An efficient numerical method for intergranular fission gas evolution under transient with piecewise boundary resolution  

NASA Astrophysics Data System (ADS)

It is theoretically found that the boundary gas concentration could decline under transient due to the intergranular resolution. An efficient numerical method is first developed for the case before saturation. After the saturation of the grain boundary, a rate-release equation is applied not only to ensure no return of the gas released, but also ensure the gas conservation in the grain boundary. Accordingly, an efficient predict-correct algorithm for the intergranular gas concentration is invented for arbitrary power transient. Our numerical method has been validated by the analytical solution and the finite element solution. It demonstrates both high efficiency as well as high accuracy.

Cui, Yi; Ding, Shurong; Huo, Yongzhong; Wang, Canglong; Yang, Lei

2013-11-01

346

Techniques for Multidimensional Measurements and Analysis of Yield of Nuclear Products Fission by Resonance Neutrons.  

National Technical Information Service (NTIS)

Using IBR-30 neutron pulse source ensuring on 57 m flight path 70 ns/m time resolution the gamma-ray spectroscopy along with the spectroscopy of neutrons following fission by resonance neutrons is carried out. An ionization chamber containing 10 g of uran...

S. A. Antonov A. A. Bogdzel' N. A. Gundorin

1985-01-01

347

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

348

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

349

EXPERIMENTAL STUDY OF THE ADSORPTION OF FISSION PRODUCTS ON ACTIVATED CARBON  

Microsoft Academic Search

A study was made of the adsorption of fission produots in solution on ; activated carbon as affected by pH. Galoride solutions of Y⁹¹, Ce¹⁴⁴-; Pr¹⁴⁴, and Cs¹³⁷ were buffered a t various pH levels and activated ; carbon put in. The use of buffers also made it possible to stady foreign-ion ; effects. The results are tabuluted and discussed

E. Spode; E. Weber

1958-01-01

350

Uncertainty Propagation of Fission Product Yield Data in Spent Fuel Inventory Calculations  

NASA Astrophysics Data System (ADS)

The effects of correlations between uncertainties of independent yields are considered to propose a method of including covariance terms within uncertainty propagation for spent fuel inventory calculations and a method outlined to achieve this. The use of the "Total Monte-Carlo" technique for such calculations are investigated for a simple decay example and for the case of a fission pulse calculation and the results discussed.

Mills, R. W.

2014-04-01

351

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

352

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 (235,238)U and (239)Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es; Fm; and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on (239)Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [H.

Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

2011-01-01

353

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

354

Large radio-frequency gas catchers and the production of radioactive nuclear beams  

NASA Astrophysics Data System (ADS)

Gas catchers provide a means to transform radioactive recoils from various production mechanisms into low-energy beams of good ion optical properties. Recent developments with large radio-frequency gas catchers have pushed back purity and space-charge limitations in this technology to the point that it can now be used reliably for producing radioactive beams intense enough for various secondary experiments to be possible. The basic technology available and the current demonstrated capabilities are presented in the following. A number of examples of such systems currently under commissioning/construction/design at ANL to produce beams from fusion-evaporation, fission, deep-inelastic and fragmentation reaction products will also be presented together with the specific challenges to each approach and the chosen solutions.

Savard, Guy

2011-09-01

355

Influence of protein fermentation on gas production profiles  

Microsoft Academic Search

With modern equipment, accurate gas-production profiles can be obtained reflecting the organic-matter fermentation in rumen fluid. Although the gas production caused by fermentation of carbohydrates is well understood and described, ignoring the influence of protein fermentation may lead to misinterpretation of the gas-production data. Gas-production profiles, from grass samples differing in growing days, and hence in protein content, showed an

John W Cone; Anthonie H van Gelder

1999-01-01

356

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

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

357

Analytical study on deformation and fission gas behavior of metallic fast reactor fuel  

Microsoft Academic Search

In order to analytically investigate irradiation behavior of metallic fast reactor fuels, the authors have developed the ALFUS (ALoyed Fuel Unified Simulator) code. The ALFUS can mechanistically simulate gas release and deformation behavior of the uranium-zirconium alloy fuel. The stress-strain analysis model into which anisotropic strain due to cavitation at the grain and\\/or phase boundary in the a-uranium phase has

Takanari Ogata; Motoyasu Kinoshita; Hiroaki Saito; Takeshi Yokoo

1996-01-01

358

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.

359

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

NASA Astrophysics Data System (ADS)

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

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

2014-08-01

360

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

361

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

NASA Astrophysics Data System (ADS)

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

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

2011-05-01

362

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

363

Development of a quantum molecular dynamic (QMD) model to describe fission and fragment production.  

PubMed

In this paper is presented the development of a QMD model for the description of the spallation reaction at energies from a few MeV to a few hundred MeV. The QMD model is developed using a new evaporation-fission model, the generalised evaporation model (GEM2). The spectrum of particles and residual nuclide mass and charge distributions in reactions of protons and neutrons with heavy targets (238U, 208Pb, 207Pb and 2206Pb) has been calculated using the QMD+GEM2 model. PMID:16604704

Polanski, A; Petrochenkov, S; Uzhinsky, V

2005-01-01

364

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

365

Comet Encke - Gas production and lightcurve  

NASA Technical Reports Server (NTRS)

A comprehensive set of observations, both from the ground and with the IUE, was planned for the 1984 apparition of Comet Encke. The observations were intended to confirm the behavior seen in 1980 and to study the behavior of the comet after perihelion. The results of the observations indicate that all the measured trace species display an asymmetry around the perihelion that is consistent with the visual light curve (VLC). But the total gas production as monitored by OH (the dominant species) displays a behavior that has no relation to the VLC.

Ahearn, M. F.; Birch, P. V.; Feldman, P. D.; Millis, R. L.

1985-01-01

366

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

SciTech Connect

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

Zhang, C. T.

1998-07-29

367

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

Microsoft Academic Search

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

Di Yun

2010-01-01

368

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

369

Gas production of three brands of ceftazidime.  

PubMed

Two sodium carbonate formulations of ceftazidime (Tazidime and Tazicef) and a new arginine formulation (Ceptaz) were evaluated for gas production and bubble formation within the drug reservoir and extension tubing of a portable infusion pump during a 24-hour delivery cycle. Triplicate samples of each brand of ceftazidime were studied under identical conditions. All formulations were constituted and diluted with sterile water for injection to a concentration of approximately 33 mg/mL, drawn into syringes, and expelled into infusion-pump drug reservoirs. Triplicate samples of degassed Tazidime and Tazicef were evaluated in the same manner. In one set of triplicate experiments, reservoirs for each formulation were attached to portable infusion pumps immediately after filling at room (23 degrees C) temperature and were programmed to deliver 25 mL over one hour every eight hours for a 24-hour delivery cycle. In a second experiment, reservoirs containing triplicate samples of each product were refrigerated (3 degrees C) for 24 hours before they were attached to the pumps for dose delivery. Visual observations were made for all pumping devices. In addition, multiple vials of each formulation were constituted, and the headspace pressure of the various formulations was monitored to compare the pressure build-up due to carbon dioxide. The presence of carbon dioxide was confirmed by gas chromatography. Pressure build-up due to carbon dioxide formation occurred in the ceftazidime sodium carbonate vials only. The sodium carbonate formulations required degrassing to reduce gas and bubble formation to a manageable level after constitution. Additionally, drug was lost because of spewing of some samples during withdrawal from the vial.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:1910261

Stiles, M L; Allen, L V; Fox, J L

1991-08-01

370

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

SciTech Connect

The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

Myers, B.F.

1995-09-01

371

Understanding the Basics of Gas Exploration and Production  

NSDL National Science Digital Library

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

Albert, Eric K.

2012-11-28

372

First-principles study of fission product (Xe, Cs, Sr) incorporation and segregation in alkaline earth metal oxides, HfO2, and MgO-HfO2 interface  

Microsoft Academic Search

In order to close the nuclear fuel cycle, advanced concepts for separating out fission products are necessary. One approach is to use a dispersion fuel form in which a fissile core is surrounded by an inert matrix that captures and immobilizes the fission products from the core. If this inert matrix can be easily separated from the fuel, via e.g.

Xiang-yang Liu; Blas P Uberuaga; Kurt E Sickafus

2008-01-01

373

First-principles study of fission product (Xe, Cs, Sr) incorporation and segregation in alkaline earth metal oxides, HfO2, and the MgO-HfO2 interface  

Microsoft Academic Search

In order to close the nuclear fuel cycle, advanced concepts for separating out fission products are necessary. One approach is to use a dispersion fuel form in which a fissile core is surrounded by an inert matrix that captures and immobilizes the fission products from the core. If this inert matrix can be easily separated from the fuel, via e.g.

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

2009-01-01

374

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 100h irradiation on a reactor with power of ~630W 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; Svadlenkov, Marie; Mil?k, Jn; Rypar, Vojt?ch; Koleka, Michal

2014-07-01

375

Isomers in Fission Fragments  

NASA Astrophysics Data System (ADS)

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 248Cm and 252Cf and in thermal neutron-induced fission of 233U, 235U and 241Pu 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.; Kster, U.; Krempel, J.; Materna, Th.; Mutti, P.; Soldner, T.; Genevey, J.; Pinston, J. A.; Simpson, G.; RzaCa-Urban, T.; Z?omaniec, A.; ?ukasiewicz, M.; Sieja, K.; Nowacki, F.; Dorvaux, O.; Gall, B. J. P.; Roux, B.; Dare, J. A.; Durell, J. L.; Smith, A. G.; Varley, B. J.; Tsekhanovich, I.; Jolie, J.; Linnemann, A.; Scherillo, A.; Orlandi, R.; Smith, J. F.; Ahmad, I.

2009-01-01

376

New integrated scheme of the closed gas-turbine cycle with synthesis gas production  

Microsoft Academic Search

New integrated scheme of the closed gas-turbine cycle with synthesis gas production was proposed. A comparative exergy analysis of the traditional gas-fired power generation cycle and proposed integrated gas-turbine cycle with synthesis gas production was carried out. The exergy losses in compressors and turbines are evaluated by using intrinsic coefficients. It has been shown that the integration of power generation

Michael S Granovskii; Mikhail S Safonov

2003-01-01

377

HTGR fuels and core development program. Quarterly progress report for the period ending May 31, 1976. [Graphite and fuels irradiation; fission product release  

SciTech Connect

The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and the data are presented in tables, graphs, and photographs.

Not Available

1976-06-30

378

Radioecological Studies in Marine Environment. A Study of the Concentration of Mixed Fission Products in Greek Sea Waters and of sup 137 Cs in Fish and Sea Plants.  

National Technical Information Service (NTIS)

The concentrations of mixed fission products in sea water and of sup 137 Cs in fish and sea plants are different from samples collected from different sampling areas. This difference is more remarkable the year where the level of the worldwide fall-out is...

S. Danali-Cotsaki H. Florou-Gazi

1982-01-01

379

Opredelenie ehnergii vozbuzhdeniya produktov spontannogo deleniya tyazhelykh yader v ramkakh statisticheskoj modeli. (The excitation energy description of heavy nuclei spontaneous fission products within the framework of statistical model).  

National Technical Information Service (NTIS)

Energies of excitation of spontaneous fission products (sup 232)Th, (sup 235,238)U, (sup 237)Np, (sup 239)Pu, (sup 241)Am and (sup 249)Cf are determined in the framework of the statistical model, in which equilibrium deformation of fragments are taken int...

A. I. Lendel T. I. Marinets D. I. Sikora E. I. Charnovich

1988-01-01

380

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

381

Analise de transmutacao considerando o tratamento explicito dos produtos de fissao num sistema acoplado composto pelos codigos Hammer-Technion e Cinder-2. (Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system).  

National Technical Information Service (NTIS)

This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECH...

A. Y. Abe

1989-01-01

382

Arrangement and method for the production of liquid natural gas  

SciTech Connect

An arrangement and a method for the increase in the production of liquid natural gas and the conservation of energy and reduction of flash gas in a liquid natural gas manufacturing installation and, more particularly, the reduction in the quantity of formed flash gas through the novel utilization of a hydraulic expander in the installation for extracting work from the flow of liquid natural gas prior to flashing thereof.

Brundige, V.L. Jr.

1984-06-26

383

Evaluation of Hot-Brine Stimulation Technique for Gas Production From Natural Gas Hydrates  

Microsoft Academic Search

Thermally efficient production of natural gas can be accomplished by the use of hot brine to dissociate solid gas hydrate deposits in the earth. The advantages of brine stimulation over steam or hot-water injection are lower energy requirements for reservoir heating and hydrate dissociation, reduced heat losses, higher gas production, and improved thermal efficiency. In addition, the problems of blockage

Vidyadhar Kamath; Sanjay Godbole

1987-01-01

384

Coal gasification process. [improvement by adding coal and clean recycle gas to the product gas  

Microsoft Academic Search

An improvement in the Koppers--Totzek coal gasification system comprises the step of adding cool and clean recycle gas to the product gas as it leaves the gasifier unit, thereby eliminating the use of water sprays to quench the product gas.

Hess

1976-01-01

385

Influence of the incident particle energy on the fission product mass distribution.  

SciTech Connect

For {sup 238}U targets and the five elements considered here, the best yields of neutron-rich isotopes are obtained from neutrons in the 2-20 MeV range. High energy beams of neutrons, protons, and deuterons have comparable integral yields per element to neutrons below 20 MeV, but the distributions are peaked at lower neutron numbers. This is presumably due to a higher neutron multiplicity in the pre-equilibrium stage and/or the compound nucleus/fission stage. For {sup 235}U targets there are high yields predicted especially for thermal neutrons, and also for the fast neutron spectrum. For the high energy neutrons, protons, and deuterons {sup 235}U has no advantage over {sup 238}U. A detailed comparison of the relative advantages of {sup 235}U and {sup 238}U for radioactive beam applications is beyond the scope of this study and will be addressed in the future. The present work is the first step of a more detailed analysis of various possible one- and two-step target geometry calculated with the LAHET code system. It is intended to serve as a guide in choosing geometry and beams for future studies. It is desirable to extend this study to higher beam energies, e.g. 200 to 1000 MeV, but at this time there is very little data against which to benchmark the analysis. Additional data would also permit comparisons of isotope yields beyond the tails of the distributions presented here, to even more neutron rich isotopes.

Gomes, I. C.

1998-08-26

386

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

387

Gas production from thermal decomposition of explosives: Assessing the thermal stabilities of energetic materials from gas production data  

Microsoft Academic Search

The gas formation associated with the thermal decompositions of nineteen energetic materials was determined at three temperatures (120C, 220C and 320C). Although there was considerable variability within classes, among the largest producers of gas were the nitrate esters. PETN (pentaerythritol nitrate) generated about 6.3mole gas per mole, while nitrocellulose, produced almost no gas. Second in gas production were the nitramines,

J. C. Oxley; J. L. Smith; E. Rogers; X. X. Dong

2000-01-01

388

Fission gas release from high burnup ThO/sub 2/ and ThO/sub 2/-UO/sub 2/ fuels irradiated at low temperature. (LWBR/AWBA development program). [LWBR, below 2700/sup 0/F  

SciTech Connect

Fission gas release data are presented for five fuel rods irradiated at low fuel temperature (below 2700/sup 0/F) with burnups up to 90,000 MWD/MTM. Four of these rods contained ThO/sub 2/-UO/sub 2/ (33.6 weight percent UO/sub 2/) fuel pellets; the fifth rod contained ThO/sub 2/ pellets. These data supplement fission gas release information previously reported for 54 rods containing ThO/sub 2/-UO/sub 2/ and ThO/sub 2/ fuel, some of which experienced fuel temperaures up to 5000/sup 0/F and burnups to 56,000 MWD/MTM. These new data suggest that at burnups exceeding about 80,000 MWD/MTM a sharp increase in fission gas release occurs, possibly caused by microstructural changes in the fuel. This is similar to the behavior of UO/sub 2/ except that the increase occurs in UO/sub 2/ at lower burnup (approximately 40,000 MWD/MTM). The fission gas release calculational model previously reported has been modified to account for the observed increase in the low temperature component. The revised model provides a good best estimate of all the fission gas release data.

Giovengo, J.F.; Goldberg, I.; Sphar, C.D.

1982-05-01

389

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

390

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

391

Gas Hearth Products Market Fact Base. Topical Report, January 1996.  

National Technical Information Service (NTIS)

The Gas Hearth Products Market Fact Base is an analysis of the U.S. gas log and fireplace markets. The study was undertaken to: determine current usage of and attitudes about fireplaces; identify barriers to acceptance of gas logs and fireplaces; determin...

1996-01-01

392

Modeling landfill gas production and movement: Principal landfill gases  

Microsoft Academic Search

A landfill gas generation and movement model is presented in this dissertation. The model is based on solution of a Darcy's law formulation of single component fluid flow in porous media in three dimensions, using a finite element technique. The effects of varying gas production rates, material porosities, and landfill covers, liners, and gas extraction wells are incorporated in the

1991-01-01

393

17 years of gas production from coal. [SASOL  

Microsoft Academic Search

The plant of South African Coal Oil and Gas Corp., Ltd., at Sasolburg was expanded to produce gas for ammonia synthesis and a 500 Btu industrial gas in addition to synthetic liquid hydrocarbons. The plant uses Lurgi coal gasification, Lurgi Rectisol, and Lurgi Phenosolvan processes. Present production of 219 million cu ft\\/day is provided by 11 of the 13 installed

Hoogendom

1972-01-01

394

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

395

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

396

Hydraulic Fracturing and Shale Gas Production: Technology, Impacts, and Policy.  

National Technical Information Service (NTIS)

Hydraulic fracturing is a key technique that has enabled the economic production of natural gas from shale deposits, or plays. The development of large-scale shale gas production is changing the U.S. energy market, generating expanded interest in the usag...

A. Burnham C. Clark C. Harto R. Horner

2012-01-01

397

Gas Production from Hydrate-Bearing Sediments - Emergent Phenomena -  

Microsoft Academic Search

Even a small fraction of fine particles can have a significant effect on gas production from hydrate-bearing sediments and sediment stability. Experiments were conducted to investigate the role of fine particles on gas production using a soil chamber that allows for the application of an effective stress to the sediment. This chamber was instrumented to monitor shear-wave velocity, temperature, pressure,

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

2012-01-01

398

Gas production and migration in landfills and geological materials  

NASA Astrophysics Data System (ADS)

Landfill gas, originating from the anaerobic biodegradation of the organic content of waste, consists mainly of methane and carbon dioxide, with traces of volatile organic compounds. Pressure, concentration and temperature gradients that develop within the landfill result in gas emissions to the atmosphere and in lateral migration through the surrounding soils. Environmental and safety issues associated with the landfill gas require control of off-site gas migration. The numerical model TOUGH2-LGM (Transport of Unsaturated Groundwater and Heat-Landfill Gas Migration) has been developed to simulate landfill gas production and migration processes within and beyond landfill boundaries. The model is derived from the general non-isothermal multiphase flow simulator TOUGH2, to which a new equation of state module is added. It simulates the migration of five components in partially saturated media: four fluid components (water, atmospheric air, methane and carbon dioxide) and one energy component (heat). The four fluid components are present in both the gas and liquid phases. The model incorporates gas-liquid partitioning of all fluid components by means of dissolution and volatilization. In addition to advection in the gas and liquid phase, multi-component diffusion is simulated in the gas phase. The landfill gas production rate is proportional to the organic substrate and is modeled as an exponentially decreasing function of time. The model is applied to the Montreal's CESM landfill site, which is located in a former limestone rock quarry. Existing data were used to characterize hydraulic properties of the waste and the limestone. Gas recovery data at the site were used to define the gas production model. Simulations in one and two dimensions are presented to investigate gas production and migration in the landfill, and in the surrounding limestone. The effects of a gas recovery well and landfill cover on gas migration are also discussed.

Nastev, Miroslav; Therrien, Ren; Lefebvre, Ren; Glinas, Pierre

2001-11-01

399

Modeling of Gas Production from Unconfined Hydrate Reservoirs  

NASA Astrophysics Data System (ADS)

Description of material: Large quantities of natural gas hydrates are present in marine sediments along the coastlines of many countries as well as in the arctic region. The production of gas from these naturally occurring gas hydrates is difficult due to complexity of thermodynamics and fluid flow involved in the process. This research is aimed at assessing production of natural gas from unconfined marine deposits of methane gas hydrates. An implicit, multiphase, multi-component, thermal, 3D simulator is used which can simulate formation and dissociation of hydrates in porous media in both equilibrium and kinetic modes. Three components (hydrate, methane and water) and four phases (hydrate, gas, aqueous-phase and ice) are considered. In this work we simulate depressurization and warm water flooding for gas production from hydrates in reservoirs underlain by an unconfined aquifer layer. Water flooding has been studied as a function of injection temperature, injection pressure, production pressure and degree of un-confinement. Application: In order to produce gas from hydrates economically, efficient production techniques must be developed. Experiments on hydrates are difficult to perform; feasibility of production can be found from simulations. Hydrate reservoirs associated with unconfined aquifer beneath are not uncommon. The determination of injection and production conditions for these reservoirs through simulation will help in designing the effective production techniques. Results and discussion: For the unconfined reservoirs associated with large aquifers the production by depressurization is inefficient. Water from the aquifer maintains the pressure in the reservoir except in the near-well regions. Warm water flooding is very effective in hydrate dissociation. Sensitivity of gas production to injection and production well conditions and degree of un-confinement has been studied. Significant new contribution: Production strategy for unconfined hydrate reservoirs.

Phirani, J.; Mohanty, K.; Hirasaki, G.

2008-12-01

400

Biomass pyrolysis\\/gasification for product gas production: the overall investigation of parametric effects  

Microsoft Academic Search

The conventional biomass pyrolysis\\/gasification process for production of medium heating value gas for industrial or civil applications faces two disadvantages, i.e. low gas productivity and the accompanying corrosion of downstream equipment caused by the high content of tar vapour contained in the gas phase. The objective of this paper is to overcome these disadvantages, and therefore, the effects of the

G Chen; J Andries; Z Luo; H Spliethoff

2003-01-01

401

METHOD OF SEPARATING FISSION PRODUCTS FROM FUSED BISMUTH-CONTAINING URANIUM  

Microsoft Academic Search

A process is described for removing metal selectively from liquid metal ; compositions. The method effects separation of flssion product metals ; selectively from dilute solution in fused bismuth, which contains uraniunn in ; solution without removal of more than 1% of the uranium. The process comprises ; contacting the fused bismuth with a fused salt composition consisting of sodium,

Wiswall

1958-01-01

402

Simultaneous separation of cesium and strontium from spent nuclear fuel using the fission-product extraction process  

SciTech Connect

The Fission-Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Global Nuclear Energy Partnership (GNEP) for the simultaneous separation of cesium and strontium from spent LWR fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository and, when combined with the separation of Am and Cm, could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly-specific extractants: 4,4',(5')-di-(t-butyl-dicyclohexano)- 18-crown-6 (DtBuCH18C6) and calix[4]arene-bis-(t-octyl-benzo-crown-6 ) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium, and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with simulated and actual spent-nuclear-fuel feed solution in centrifugal contactors are detailed. Removal efficiencies, co-extraction of metals, and process hydrodynamic performance ar e discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel. Recent advances in the evaluation of alternative calixarenes with increased solubility and stability are also detailed. (authors)

Law, J.D.; Peterman, D.R.; Riddle, C.L.; Meikrantz, D.A.; Todd, T.A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3870 (United States)

2008-07-01

403

On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel  

NASA Astrophysics Data System (ADS)

Radiation induced oxidative dissolution of UO 2 is a key process for the safety assessment of future geological repositories for spent nuclear fuel. This process is expected to govern the rate of radionuclide release to the biosphere. In this work, we have studied the catalytic effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel. The experimental studies were performed using UO 2 pellets containing 0%, 0.1%, 1% and 3% Pd as a model for spent nuclear fuel. H 2O 2 was used as a model for radiolytical oxidants (previous studies have shown that H 2O 2 is the most important oxidant in such systems). The pellets were immersed in aqueous solution containing H 2O 2 and HCO3- and the consumption of H 2O 2 and the dissolution of uranium were analyzed as a function of H 2 pressure (0-40 bar). The noble metal inclusions were found to catalyze oxidation of UO 2 as well as reduction of surface bound oxidized UO 2 by H 2. In both cases the rate of the process increases with increasing Pd content. The reduction process was found to be close to diffusion controlled. This process can fully account for the inhibiting effect of H 2 observed in several studies on spent nuclear fuel dissolution.

Trummer, Martin; Nilsson, Sara; Jonsson, Mats

2008-08-01

404

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

USGS Publications Warehouse

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

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

2012-01-01

405

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

406

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

SciTech Connect

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

Woodley, R.E.

1986-06-01

407

How technology and price affect US tight gas potential. Part 1. Technology of tight gas production  

Microsoft Academic Search

The tight gas resource in the US currently is estimated at 900 tcf, of which 600 tcf is considered technically recoverable. This gas is found in basins that cover a prospective area of one million square miles (one million sections). Of these, ca 120,000 sections are potentially productive. The tight gas picture is composed of many different and often complex

R. W. Jr. Veatch; O. Baker

1983-01-01

408

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

Microsoft Academic Search

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

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

2010-01-01

409

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

SciTech Connect

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

Hanan, N. A.

1998-10-14

410

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

411

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

SciTech Connect

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

Marra, J.; Billings, A.

2009-06-24

412

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

SciTech Connect

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

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

2010-02-26

413

A repository released-dose model for the evaluation of long-lived fission product transmutation effectiveness  

SciTech Connect

A methodology has been developed to quantify the total integrated dose due to a radionuclide species i emplaced in a geologic repository; the focus is on the seven long-lived fission products (LLFPs). The methodology assumes continuous exposure water contaminated with species i at the accessible environment (i.e., just beyond the geologic barrier afforded by the geologic repository). The dose integration is performed out to a reference post-release time. The integrated dose is a function of the total initial inventory of radionuclide i the repository, the time at which complete and instantaneous failure of the engineered barrier (e.g., waste canister) in, a geologic repository occurs, the fractional dissolution rate (from waste solid form) of radionuclide i in ground water, the ground water travel time to the accessible environment, the retardation factor (sorption on the geologic media) for radionuclide i, the time after radionuclide begins to enter the biosphere. In order to assess relative dose, the ratio of total integrated dose to that for a reference LLFP species j (e.g., {sup 99}Tc) was defined. This ratio is a measure of the relative benefit of transmutation of other LLFPs compared to {sup 99}Tc. This methodology was further developed in order to quantify the integrated dose reduction per neutron utilized for LLFP transmutation in accelerator-driven transmutation technologies (ADTT). This measure of effectiveness is a function of the integrated dose due to LLFP species i, the number of total captures in LLFP species i chain per LLFP nuclide fed to the chain at equilibrium, and the number of total captures in related transmutation product (TP) chains per capture in the LLFP species i chain. To assess relative transmutation effectiveness, the ratio of integrated dose reduction per neutron utilization to that for a reference LLFP species j (e.g., {sup 99}Tc) was defined. This relative measure of effectiveness was evaluated LLFP transmutation strategy.

Davidson, J.W.

1995-07-01

414

Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts  

NASA Astrophysics Data System (ADS)

The removal of fission product elements from molten salt wastes arising from pyrochemical reprocessing of spent nuclear fuels has been investigated. The experiments were conducted in LiCl-KCl eutectic at 550 C and NaCl-KCl equimolar mixture at 750 C. The behavior of the following individual elements was investigated: Cs, Mg, Sr, Ba, lanthanides (La to Dy), Zr, Cr, Mo, Mn, Re (to simulate Tc), Fe, Ru, Ni, Cd, Bi and Te. Lithium and sodium phosphates were used as precipitants. The efficiency of the process and the composition of the solid phases formed depend on the melt composition. The distribution coefficients of these elements between chloride melts and precipitates were determined. Some volatile chlorides were produced and rhenium metal was formed by disproportionation. Lithium-free melts favor formation of double phosphates. Some experiments in melts containing several added fission product elements were also conducted to study possible co-precipitation reactions. Rare earth elements and zirconium can be removed from both the systems studied, but alkaline earth metal fission product elements (Sr and Ba) form precipitates only in NaCl-KCl based melts. Essentially the reverse behavior was found with magnesium. Some metals form oxide rather than phosphate precipitates and the behavior of certain elements is solvent dependent. Caesium cannot be removed completely from chloride melts by a phosphate precipitation technique.

Volkovich, Vladimir A.; Griffiths, Trevor R.; Thied, Robert C.

2003-11-01

415

The Effect of Oral ?-Galactosidase on Intestinal Gas Production and Gas-Related Symptoms  

Microsoft Academic Search

Bloating, abdominal distention, and flatulence represent very frequent complaints in functional disorders but their pathophysiology\\u000a and treatment are largely unknown. Patients frequently associate these symptoms with excessive intestinal gas and the reduction\\u000a of gas production may represent an effective strategy. The aim was to evaluate the effect of ?-galactosidase administration,\\u000a in a randomized double-blind placebo-controlled protocol, on intestinal gas production

Michele Di Stefano; Emanuela Miceli; Samantha Gotti; Antonio Missanelli; Samanta Mazzocchi; Gino Roberto Corazza

2007-01-01

416

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

417

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

SciTech Connect

During the previous funding cycle for this project, we investigated the electrochemistry of Cs(I) in air and moisture-stable ionic liquids both with and without the addition of BOBCalixC6. These investigations revealed that the electrochemical windows of the dialkylimidazolium bis[(trifluoromethyl)sulfonyl]imide ionic liquids do not permit the direct electrochemical reduction of Cs(I), even when Hg electrodes are employed, because these organic cations are reduced at less negative potentials than Cs(I). However, Cs(I) coordinated by BOBCalixC6 can be electrolytically reduced to Cs(Hg) in tetraalkylammonium-based room-temperature ionic liquids such as tri-1-butylmethylammonium bis[(trifluoromethyl)sulfonyl]imide (Bu3MeN+Tf2N-) at Hg electrodes. Because this reduction process does not harm either the ionic liquid or the macrocycle, it is a promising method for recycling the cesium extraction system. The previous studies mentioned above were carried out under an inert atmosphere, i.e., in the absence of H2O and O2. However, it may not be economically feasible or even possible to carry out the recycling process in the absence of these contaminants during large-scale processing of aqueous tank waste. Thus, as described in our proposal, we have begun an investigation of the electrochemical recovery of Cs from the Bu3MeN+Tf2N- + BOBCalixC6 extraction system in an air atmosphere containing various amounts of water and oxygen. Our recent preliminary results were very surprising because they indicated that the electrochemical extraction process is relatively insensitive to the presence of small amounts of moisture even when the moisture content of the ionic liquid approaches 1000 ppm. Furthermore, we have found that the ''wet'' ionic liquid can be easily dehydrated under reduced pressure or by sparging with dry nitrogen gas without the need for heat or any other specialized treatment.

Hussey, Charles L.

2005-06-01

418

Structure and shale gas production patterns from eastern Kentucky field  

SciTech Connect

Computer-derived subsurface structure, isopach, and gas-flow maps, based on 4000 drillers logs, have been generated for eastern Kentucky under a project sponsored by the Gas Research Institute. Structure maps show low-relief flextures related to basement structure. Some structures have been mapped at the surface, others have not. Highest final open-flow (fof) of shale gas from wells in Martin County follow a structural low between (basement) anticlines. From there, elevated gas flows (fof) extend westward along the Warfield monocline to Floyd County where the high flow (fof) trend extends southward along the Floyd County channel. In Knott County, the number of wells with high gas flow (fof) decreases abruptly. The center of highest gas flow (fof) in Floyd County spreads eastward to Pike County, forming a triangular shaped area of high production (fof). The center of highest gas flow (fof) is in an area where possible (basement) structure trends intersect and where low-relief surface folds (probably detached structure) were mapped and shown on the 1922 version of the Floyd County structure map. Modern regional maps, based on geophysical logs from widely spaced wells, do not define the low-relief structures that have been useful in predicting gas flow trends. Detailed maps based on drillers logs can be misleading unless carefully edited. Comparative analysis of high gas flows (fof) and 10-year cumulative production figures in a small area confirms that there is a relationship between gas flow (fof) values and long-term cumulative production.

Shumaker, R.C.

1987-09-01

419

Fission Products in National Atmospheric Deposition Program--Wet Deposition Samples Prior to and Following the Fukushima Dai-Ichi Nuclear Power Plant Incident, March 8-April 5, 2011.  

National Technical Information Service (NTIS)

Radioactive isotopes I-131, Cs-134, or Cs-137, products of uranium fission, were measured at approximately 20 percent of 167 sampled National Atmospheric Deposition Program monitoring sites in North America (primarily in the contiguous United States and A...

C. M. Lehmann D. A. Gay G. A. Wetherbee M. A. Nilles T. M. Debey

2012-01-01

420

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

SciTech Connect

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

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

1980-01-01

421

Gas hearth products market fact base. Topical report, January 1996  

SciTech Connect

The Gas Hearth Products Market Fact Base is an analysis of the U.S. gas log and fireplace markets. The study was undertaken to: determine current usage of and attitudes about fireplaces; identify barriers to acceptance of gas logs and fireplaces; determine the influence of service providers, and; identify important trends that can affect the markets for gas hearth products. The market fact base is based on four studies: a market analysis synthesizing primary and secondary research reports; in-depth interviews with market influencers from across the country (architects, contractors, interior designers, fireplace retailers and installers) and industry experts from gas utilities and trade associations; focus group meetings with consumers who own or intend to buy fireplaces, gas fireplace industry professionals, and editors of fireplace-related trade magazines, and; quantitative interviews with consumers in six U.S. cities.

NONE

1996-02-01

422

Strategies for gas production from oceanic Class 3 hydrateaccumulations  

SciTech Connect

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

Moridis, George J.; Reagan, Matthew T.

2007-05-01

423

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) [Torrance, CA; Najjar, Mitri S. (Wappingers Falls, NY) [Wappingers Falls, NY

1993-01-01

424

Cathodic H2 gas production through Pd alloy membrane electrodes  

NASA Astrophysics Data System (ADS)

A rechargeable H2-NiOOH cell with hydrogen-permeable membrane electrode was tested, and its cathodic hydrogen gas production through the membrane electrode investigated. When a Pd-Pt, catalyzed electrolyte-facing surface was cathodically polarized in a concentrated KOH solution, it was found that hydrogen gas was evolved in the chamber through dissolved hydrogen atoms' penetrating of the membrane to exit at the other, palladized surface as free gas.

Shirogami, T.; Murata, K.

425

30 CFR 250.1102 - Oil and gas production rates.  

Code of Federal Regulations, 2010 CFR

...1102 Oil and gas production rates. (a) MER. (1) The lessee shall submit a proposed MER for each producing sensitive reservoir on... (2) The lessee may propose to revise an MER by submitting Form MMS-127 with...

2009-07-01

426

Production of a pulseable fission-like neutron flux using a monoenergetic 14 MeV neutron generator and a depleted uranium reflector  

NASA Astrophysics Data System (ADS)

The design and performance of a pulseable neutron source utilizing a D-T neutron generator and a depleted uranium reflector are presented. Approximately half the generator's 14 MeV neutron flux is used to produce a fission-like neutron spectrum similar to 252Cf. For every 14 MeV neutron entering the reflector, more than one fission-like neutron is reflected back across the surface of the reflector. Because delayed neutron production is more than two orders of magnitude below the prompt neutron production, the source takes full advantage of the generator's pulsed mode capability. Applications include all elemental characterization systems using neutron-induced gamma-ray spectroscopy. The source simu