These are representative sample records from Science.gov related to your search topic.
For comprehensive and current results, perform a real-time search at Science.gov.
1

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

2

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

SciTech Connect

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

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

2008-07-01

3

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

4

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

5

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

6

The primary charge of fission products from the thermal neutron fission of 235U  

Microsoft Academic Search

The average primary charge Zp of fission products from thermal neutron fission of 235U is determined by measuring the number of beta-decays suffered by the fission products, which are mass separated with the gas-filled mass separator of the research reactor FRJ-2. The measurement is performed with a 4pi plastic detector to avoid errors caused by conversion electrons and gamma-rays emitted

K. Sistemich; P. Armbruster; J. Eidens; E. Roeckl

1969-01-01

7

Correlation of recent fission product release data  

SciTech Connect

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

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

1989-01-01

8

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment  

Microsoft Academic Search

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

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

2010-01-01

9

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

NASA Technical Reports Server (NTRS)

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

Capo, M. A.; Mickle, R.

1975-01-01

10

Fission Product Release from SLOWPOKE-2 Reactors  

NASA Astrophysics Data System (ADS)

Increasing radiation fields at several SLOWPOKE -2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace above the reactor were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements.

Harnden, Anne M. C.

11

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

12

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  

SciTech Connect

The US Department of Energy (DOE) 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 fuel irradiation tests are planned for the Idaho National Laboratory's (INL's) advanced test reactor (ATR). The first test in this series (AGR-1) will incorporate six separate capsules irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitored in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded, high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the fission product monitoring 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.

John K. Hartwell

2007-06-01

13

Rapid separation of fresh fission products (draft)  

Microsoft Academic Search

The fission of highly eruiched uranium by thermal neutrons creates dozens of isotopic products. The Isotope and Nuclear Chemistry Group participates in programs that involve analysis of 'fiesh' fission products by beta counting following radiochemical separations. This is a laborious and time-consuming process that can take several days to generate results. Gamma spectroscopy can provide a more immediate path to

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

2003-01-01

14

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

15

Reactor power history from fission product signatures  

E-print Network

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

Sweeney, David J.

2009-05-15

16

Fission product release from irradiated LWR fuel under accident conditions  

SciTech Connect

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

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

1984-01-01

17

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

18

Rapid separation of fresh fission products (draft)  

SciTech Connect

The fission of highly eruiched uranium by thermal neutrons creates dozens of isotopic products. The Isotope and Nuclear Chemistry Group participates in programs that involve analysis of 'fiesh' fission products by beta counting following radiochemical separations. This is a laborious and time-consuming process that can take several days to generate results. Gamma spectroscopy can provide a more immediate path to isolopic activities, however short-lived, high-yield isotopes can swamp a gamma spectrum, making difficult the identification and quantification of isotopes on the wings and valley of the fission yield curve. The gamma spectrum of a sample of newly produced fission products is dominated by the many emissions of a very few high-yield isotopes. Specilkally, {sup 132}Te (3.2 d), its daughter, {sup 132}I(2 .28 h), {sup 140}Ba (12.75 d), and its daughter {sup 140}La (1.68 d) emit at least 18 gamma rays above 100 keV that are greater than 5% abundance. Additionally, the 1596 keV emission fiom I4'La imposes a Compton background that hinders the detection of isotopes that are neither subject to matrix dependent fractionation nor gaseous or volatile recursors. Some of these isotopes of interest are {sup 111}Ag, {sup 115}Cd, and the rare earths, {sup 153}Sm, {sup 154}Eu, {sup 156}Eu, and {sup 160}Tb. C-INC has performed an HEU irradiation and also 'cold' carrier analyses by ICP-AES to determine methods for rapid and reliable separations that may be used to detect and quantify low-yield fission products by gamma spectroscopy. Results and progress will be presented.

Dry, D. E. (Donald E.); Bauer, E. (Eve); Petersen, L. A. (Lisa A.)

2003-01-01

19

Mass distribution of fission products following photofission of uranium-238  

Microsoft Academic Search

The mass-yield distribution of fission products following photofission ; of ²³⁸U using bremsstrahlung energies of 22, 24, and 28 MeV were measured ; by radiochemically isolating the fission products belonging to 24 mass chains. ; The absolute activities of these nuclides were determined by BETA - and gamma ; counting techniques, and the cumulative fission yields were calculated relative ;

D. Swindle; R. Wright; K. Takahashi; W. H. Rivera; J. L. Meason

1973-01-01

20

Time dependent particle emission from fission products  

SciTech Connect

Decay heating following nuclear fission is an important factor in the design of nuclear facilities; impacting a variety of aspects ranging from cooling requirements to shielding design. Calculations of decay heat, often assumed to be a simple product of activity and average decay product energy, are complicated by the so called 'pandemonium effect'. Elucidated in the 1970's this complication arises from beta-decays feeding high-energy nuclear levels; redistributing the available energy between betas and gammas. Increased interest in improving the theoretical predictions of decay probabilities has been, in part, motivated by the recent experimental effort utilizing the Total Absorption Gamma-ray Spectrometer (TAGS) to determine individual beta-decay transition probabilities to individual nuclear levels. Accurate predictions of decay heating require a detailed understanding of these transition probabilities, accurate representation of particle decays as well as reliable predictions of temporal inventories from fissioning systems. We will discuss a recent LANL effort to provide a time dependent study of particle emission from fission products through a combination of Quasiparticle Random Phase Approximation (QRPA) predictions of beta-decay probabilities, statistical Hauser-Feshbach techniques to obtain particle and gamma-ray emissions in statistical Hauser-Feshbach and the nuclear inventory code, CINDER.

Holloway, Shannon T [Los Alamos National Laboratory; Kawano, Toshihiko [Los Alamos National Laboratory; Moller, Peter [Los Alamos National Laboratory

2010-01-01

21

Fission and Nuclear Liquid-Gas Phase Transition  

E-print Network

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

E. A. Cherepanov; V. A. Karnaukhov

2007-03-30

22

FFTF (Fast Flux Test Facility) Fission Gas Monitor Computer System  

Microsoft Academic Search

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled, fast neutron test reactor located on the Hanford Site. A dual computer system has been developed to monitor the reactor cover gas to detect and characterize any fuel or test pin fission gas releases. The system acquires gamma spectra data, identifies isotopes, calculates specific isotope and overall cover gas activity, presents

J. A. Hubbard; G. T. Taylor

1987-01-01

23

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

SciTech Connect

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

Rest, J.

1983-09-01

24

Energy production using fission fragment rockets  

NASA Astrophysics Data System (ADS)

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

Chapline, G.; Matsuda, Y.

1991-08-01

25

Energy production using fission fragment rockets  

SciTech Connect

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs.

Chapline, G.; Matsuda, Y.

1991-08-01

26

Calculations on fission gas behaviour in the high burnup structure  

NASA Astrophysics Data System (ADS)

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

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

2006-05-01

27

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

28

ELSA: A simplified code for fission product release calculations  

SciTech Connect

During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species.

Manenc, H. [Institut de Protection et de Surete Nucleaire, St. Paul-lez-Durance (France); Notley, M.J. [MJF International, Oundle (United Kingdom)

1996-12-31

29

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

SciTech Connect

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

Acharya, R.; Hanson, D.

1988-05-01

30

Migration behavior of fission products in and from spherical high-temperature reactor fuel elements  

SciTech Connect

Diffusion behavior of some metallic fission products in high-temperature reactor fuel elements, which had been irradiated in an in-pile gas loop (Saphir) installed in the Pegase reactor (France), was studied. Diffusion coefficients of cesium and silver in hightemperature isotropic pyrolytic carbon and graphite matrix under in-pile conditions were obtained by analyzing the concentration profiles of the fission products in the fuel elements, which had been measured by postirradiation examination. Although ruthenium profiles were measured, analysis of the diffusion coefficients could not be carried out because of the virtually flat distributions. By comparing the concentrations of the cesium isotopes in the fuel-free zone of the elements, it was found that TUCs behaved anomalously in the graphite matrix, which was, probably, caused by activation of an undetectable amount of TTCs impurity involved in the matrix. For the extremely high concentration of these fission products, which had been observed near the surface of the element, two causes, the uranium contamination concentrating there and the trapping effect in the defects introduced by fission of the locally concentrated uranium, were considered, although these high concentrations of the fission products were neglected in the analysis. Furthermore, Transport behavior of the fission products through the gas gap from the fuel element to the graphite tube containing the elements was studied by measuring the concentration profiles in the tube. It was concluded that ruthenium transport occurred by direct fission recoil from the surface uranium contamination, whereas that of cesium, by desorption from the surface.

Fukuda, K.; Groos, E.; Rau, J.

1985-06-01

31

Fission-product yields for thermal-neutron fission of curium-243  

SciTech Connect

Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of /sup 243/Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of /sup 243/Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for /sup 243/Cm.

Breederland, D. G.

1982-01-01

32

URANIUM235 FISSION-PRODUCT PRODUCTION AS A FUNCTION OF THE THERMAL NEUTRON FLUX, IRRADIATION TIME, AND DECAY TIME. II. SUMMATIONS OF INDIVIDUAL CHAINS, ELEMENTS, AND THE RARE-GAS AND RARE-EARTH GROUPS. VOLUMES 1 AND 2  

Microsoft Academic Search

These two volumes were issued separately, but are cataloged as a unit. ; The following properties, per initial atom of U²³⁵, are tabulated for each ; fission-product chain with mass number 72 to 161: activity, gamma power, total ; power, poisoning, and gamma disintegrations per second with energies up to 1.70 ; Mev. (M.H.R.);

J. O. Blomeke; M. F. Todd

1958-01-01

33

Release of fission products from irradiated aluminide fuel at high temperature  

SciTech Connect

Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500/sup 0/C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1/sup 0/C on the plates. The next release at 585/sup 0/C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640/sup 0/C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed.

Shibata, T.; Kanda, K.; Mishima, K.

1982-01-01

34

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

E-print Network

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

Andrews, Nathan Christopher

2012-01-01

35

Analysis of Fission Products on the AGR-1 Capsule Components  

SciTech Connect

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

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

2013-03-01

36

Modeling fission product vapor transport in the Falcon facility  

SciTech Connect

An extensive database of aerosol Experiments exists and has been used for checking aerosol transport codes. Data for fission product vapor transport are harder to find. Some qualitative data are available, but the Falcon thermal gradient tube tests carried out at AEA Technology`s laboratories in Winfrith, England, mark the first serious attempt to provide a set of experiments suitable for the validation of codes that predict the transport and condensation of realistic mixtures of fission product vapors. Four of these have been analyzed to check how well the computer code VICTORIA can predict the most important phenomena. Of the four experiments studied, two are reference cases (FAL-17 and FAL-19), one is a case without boric acid (FAL-18), and the other is run in a reducing atmosphere (FAL-20). The results show that once the vapors condense onto aerosols, VICTORIA can predict their deposition rather well. The dominant mechanism is thermophoresis, and each element deposits with more or less the same deposition velocity. The behavior of the vapors is harder to interpret. Essentially, it is important to know the temperature at which each element condenses. It is clear from the measurements that this temperature changed from test to test-caused mostly by the different speciation as the composition of the carrier gas and the relative concentration of other fission products changed. Only in the test with a steam atmosphere and without boric acid was the assumption valid that most of the iodine is cesium iodide and most of the cesium is cesium hydroxide. In general, VICTORIA predicts that, with the exception of cesium, there will be less variation in the speciation-and, hence, variation in the deposition-between tests than is in fact observed. VICTORIA underpredicts the volatility of most elements, and this is partly a consequence of the ideal solution assumption and partly an overestimation of vapor/aerosol interactions.

Shepherd, I.M.; Drossinos, Y. [Joint Research Center, Ispra (Italy); Benson, C.G. [AEA Technology, Winfrith (United Kingdom)

1995-05-01

37

Fission product removal from molten salt using zeolite  

Microsoft Academic Search

Spent nuclear fuel (SNF) can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. The fuel is dissolved in molten chloride salt. Non-transuranic fission products in the molten salt are ion-exchanged into zeolite A, which is subsequently mixed with glass and consolidated. Zeolite was found to be effective in removing fission

C. Pereira; B. D. Babcock

1996-01-01

38

GAMMA SPECTRA OF GROSS FISSION PRODUCTS FROM THERMAL REACTORS  

Microsoft Academic Search

Calculations of gamma spectra from products of thermal fission of U\\/sup ; 235\\/ are presented. The fission yield values and decay data used are taken from ; literature published up to April 1958. The calculations cover seven different ; irradiation times from one day to two years and continuous cooling times from one ; day to 1000 years. The gamma

J. Prawitz; K. Low; R. Bjornerstedt

1959-01-01

39

(COMEDIE program review and fission product transport in MHTGR reactor)  

SciTech Connect

The subcontract between Martin Marietta Energy Systems, Inc., and the CEA provides for the refurbishment of the high pressure COMEDIE test loop in the SILOE reactor and a series of experiments to characterize fission product lift-off from MHTGR heat exchanger surfaces under several depressurization accident scenarios. The data will contribute to the validation of models and codes used to predict fission product transport in the MHTGR. In the meeting at CEA headquarters in Paris the program schedule and preparation for the DCAA and Quality Assurance audits were discussed. Long-range interest in expanded participation in the gas-cooled reactor technology Umbrella Agreement was also expressed by the CEA. At the CENG, in Grenoble, technical details on the loop design, fabrication components, development of test procedures, and preparation for the DOE quality assurance (QA) audit in May were discussed. After significant delays in CY 1989 it appears that good progress is being made in CY 1990 and the first major test will be initiated by December. An extensive list of agreements and commitments was generated to facilitate the coordination and planning of future work. 2 figs., 2 tabs.

Stansfield, O.M.

1990-03-15

40

Fission-product SiC reaction in HTGR fuel  

SciTech Connect

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

Montgomery, F.

1981-07-13

41

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

Microsoft Academic Search

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

J. K. Dickens; J. W. McConnell

1981-01-01

42

Characteristics of the Fission Product Cs135  

Microsoft Academic Search

Two samples of Xe135 of high intensity were prepared at the Los Alamos homogeneous pile. The decay products of the gas contained a Cs activity presumed to be Cs135. The half-life of Cs135 was found to be (2.1+\\/-0.7)106 yr. and the maximum energy of its beta--radiations 0.21 Mev. No gamma-radiations were detected. The capture cross section of Cs135 for pile

Nathan Sugarman

1949-01-01

43

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

44

Thermodynamics of fission products in UO(2 x).  

PubMed

The stabilities of selected fission products-Xe, Cs, and Sr-are investigated as a function of non-stoichiometry x in UO(2 x). In particular, density functional theory (DFT) 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(2), the DFT calculations are performed using spin polarization and with the Hubbard U term. 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(2)O is observed as a second stable phase and SrO is found to be soluble in the UO(2) matrix for all stoichiometries. These observations mirror experimentally observed phenomena. PMID:21832440

Nerikar, P V; Liu, X-Y; Uberuaga, B P; Stanek, C R; Phillpot, S R; Sinnott, S B

2009-10-28

45

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

46

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

SciTech Connect

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

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

2012-07-01

47

The Oklo natural reactor: Cumulative fission yields and retentivity of the symmetric mass region fission products  

NASA Astrophysics Data System (ADS)

Solid source mass spectrometry has been used to determine the relative cumulative fission yields of five elements in three samples of uranium ore from reactor zones in the Oklo mine site. Eighteen fission chains covering the mass range from 105 ? A ? 130 have been measured for Pd, Ag, Cd, Sn and Te. These measurements have enabled a number of nuclear parameters to be calculated including the relative proportions of 235U, 238U and 239Pu involved in the fission process. The concentration of the five elements in the Oklo samples have also been measured using the stable isotope dilution technique. These values have then been compared to the estimates of the amount of these elements produced by fission under the conditions that are appropriate to the three samples. This procedure enables the retentivity of the elements in the reactor zones to be evaluated. Our work confirms the fact that Pd and Te are retained almost in their entirety in the samples, whereas the other three elements have been partially lost from the reactor site. Almost all the Cd fission products have been lost, and more than 50% of the Ag and Sn fission-produced material has been removed.

De Laeter, J. R.; Rosman, K. J. R.; Smith, C. L.

1980-10-01

48

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

Microsoft Academic Search

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

HYEDONG JEONG; SOON HEUNG CHANG

49

Experiments and analysis of fission product release in HEU-fuelled SLOWPOKE-2 reactors  

NASA Astrophysics Data System (ADS)

Fission product activity levels have been measured using a transportable gamma ray spectroscopy system at four SLOWPOKE-2 facilities. Through an analysis of the concentrations of these radionuclides in samples of the reactor coolant and gas headspace, the rate of release from the fuel has been determined by a Savitzky-Golay method and also by a non-linear least squares method. The release rate calculation has been validated against the mainframe code SUMRT. By examining the release rates, the source of the short-lived fission products is determined to be direct recoil from exposed uranium-bearing surfaces.

Harnden-Gillis, A. C.; Bennett, L. G. I.; Lewis, B. J.

1994-07-01

50

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

51

The behavior of fission products during nuclear rocket reactor tests  

SciTech Connect

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

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

1991-01-01

52

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

53

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

54

Fission product removal from molten salt using zeolite  

SciTech Connect

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

Pereira, C.; Babcock, B.D.

1996-10-01

55

Data summary report for fission product release test VI-5  

SciTech Connect

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

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

1991-10-01

56

Benchmark experiments for fission product data (nuclear date)  

Microsoft Academic Search

Five general areas of application for nuclear cross sections have been ; identified for benchmark testing by the Cross Section Evaluation Working Group ; (CSEWG); thermal and fast reactors, shielding, dosimetry and fission product ; properties. For both thermal and fast reactors, benchmark experiments consist of ; measurements of integral reaction rates, reactivity coefficients and criticality ; in well defined,

Schenter

1975-01-01

57

Fission product scrubbing system for a nuclear reactor  

Microsoft Academic Search

A fission product scrubbing system is described for a nuclear reactor including a containment building defining a containment space for accommodating reactor components, comprising (a) means defining a water tank in the containment building; (b) a dividing wall extending into the water tank for separating the water tank into a first and a second compartment; (c) means defining a collection

1986-01-01

58

BIOLOGIC SHIELDING AGAINST GAMMA-RAYS FROM FISSION-PRODUCTS  

Microsoft Academic Search

An effective attenuation-coefficient is determined in order to calcuiate ; the biologic shielding against gamma-rays of an originai fission product mixture. ; Its demonstration by diagrams is given as a function of the irradiation time of ; the uranium, of the cooling time after irradiation, of the wall thickness, and of ; the shielding material. (auth);

1962-01-01

59

Fission fragment study of ^252Cf using the gas-filled separator, SASSYER  

NASA Astrophysics Data System (ADS)

Nuclei lying far from the valley of beta-stability have extreme neutron to proton (N/Z) ratios, and exhibit properties at variance with their more stable counterparts. The proton dripline lies closer to stability due to the repulsive Coulomb force between protons and is more easily accessible to experiment through heavy-ion reactions. Extensive studies of neutron-rich nuclei have been more difficult, as they are produced in fission and fragmentation studies at relatively low yields. Gamma-ray spectroscopy of ^252Cf fission fragments has been performed at the Wright Nuclear Structure Laboratory of Yale University. A 10-?Ci source was placed at the center of YRAST-Ball, a high-efficiency clover Ge array. Recoiling fragments were subsequently separated through the gas-filled magnetic spectrometer SASSYER, and implanted into a Si detector at the exit. Coincidences between implanted fission fragments and prompt gamma-ray emissions were used to select mass regions of the fission products. Preliminary results of this experiment and future work will be discussed. This work was supported by the U.S. DOE under Contract Nos. DE-FG02-91ER-40609 and DE-FG02-88ER-40417.

Ai, H.; Beausang, C. W.; Ressler, J. J.; Amro, H.; Caprio, M. A.; McCutchan, E. A.; Zamfir, N. V.

2003-04-01

60

RIS-M-2599 DETERMINATION OF FISSION PRODUCTS IN IRRADIATED FUEL BY X-RAY  

E-print Network

of water reactor fuel during a power transient. The mechanisms by which the volatile fission products are set free are s t i l l a matter for discussion, although fission product release from nuclear fuel has

61

FIELD STUDIES OF FISSION PRODUCT INHALATION. PART III. FISSION PRODUCT FIELD RELEASE TEST SERIES ONE (FPFRT-1)  

Microsoft Academic Search

Rats and dogs were exposed to clouds of fission products released from ; the high temperature meltdown of reactor fuel element in the field. The primary ; goal of this biological program was attainment of good lung deposition values ; which could be used in the initial step of the biological counterpart of reactor ; hazards evaluation. There was a

R. G. Thomas; R. H. Wilson

1959-01-01

62

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

63

Fission product release from uranium-aluminum alloy fuel in Slowpoke-2 reactors  

SciTech Connect

Increasing, but still low, radiation fields due to a release of fission products have been observed in the light-water-filled reactor container of SLOWPOKE-2 reactors fueled with a highly enriched uranium alloy. To investigate this phenomenon, samples of water coolant and headspace gas from the reactor container have been examined by gamma spectroscopy methods for several reactors with various burnup. A model has been developed to describe the kinetic behavior of the activity concentrations of the short-lived iodine and noble gas species in the reactor container water, and the noble gas concentrations in the reactor container head-space. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line of the fuel elements that originated at the time of fuel fabrication. The fission product release analysis is consistent with observations from an underwater visual examination of a high-burnup core and a metallographic examination of archived fuel elements.

Lewis, B.J.; Harnden-Gillis, A.C.; Bennett, L.G.I. (Royal Military Coll. of Canada, Kingston, Ontario (Canada). Dept. of Chemistry and Chemical Engineering)

1994-03-01

64

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

65

CHROMATOGRAPHIC SEPARATION OF Np FROM U, Pu AND FISSION PRODUCTS  

Microsoft Academic Search

A method of separating Np, Pu, and U by cation exchange is described. ; Neptunium(V) is eluted from the cation column by 1M nitric acid ahead of U(VI) ; and other fission products. The it is purified by ether extraction of oxidized ; hexavalent neptunium. Cation resin KU-2 and 1M nitric acid reduce hexavalent Np ; to pentavalent. (R.J.V.);

Yu. A. Zolotov; D. Nishanov

1962-01-01

66

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

67

Data summary report for fission product release test VI3  

Microsoft Academic Search

Test VI-3, the third in a series of high-temperature fission product release tests in the vertical test apparatus, was conducted in flowing steam. The test specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium, which had been irradiated to a burnup of 42 MWd\\/kg. Using an induction furnace, it was heated under simulated light-water

M. F. Osborne; R. A. Lorenz; J. L. Collins; J. R. Travis; C. S. Webster; H. K. Lee; T. Nakamura; Y.-C. Tong

1990-01-01

68

Measurements for the JASPER Program fission gas plenum experiment  

SciTech Connect

The Fission Gas Plenum Experiment was conducted at the Oak Ridge National Laboratory Tower Shielding Facility during FY 1987 to: provide data for verification of the assumptions and calculational methods used to determine the neutron leakage from the plenum, and provide an uncertainty evaluation associated with the calculations. The Tower Shielding Reactor source was modified to represent the neutron spectrum leaving a typical liquid-metal-cooled reactor core along its axis. The experimental configurations resulted from the insertion of either a homogeneous or homogeneous-heterogeneous gas plenum combination into the iris of a concrete slab, with the only variable being the thickness of the plenum. Integral neutron fluxes were measured behind each of the configurations at specified locations, and neutron spectra were obtained behind selected mockups. The experimental data are presented in both tabular and graphical form. This experiment is the second in a series of six experiments to be performed as part of a cooperative effort between the United States Department of Energy and the Japan Power Reactor and Nuclear Fuel Development Corporation. The research program is intended to provide support for the development of advanced sodium-cooled reactors.

Muckenthaler, F.J.

1987-06-01

69

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

NASA Astrophysics Data System (ADS)

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

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

2014-05-01

70

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

71

RAFT: a computer model for formation and transport of fission product aerosols in LWR primary systems  

SciTech Connect

A computer model, RAFT (Reactor Aerosol Formation and Transport), has been developed to predict the size distribution and composition of the particles (aerosols) formed from condensation of the fission product and control rod material vapors released in LWR accidents. The underlying theory of RAFT considers the processes of homogeneous and heterogeneous nucleation, aerosol agglomeration, and aerosol and vapor deposition, in conjunction with the equilibrium chemistry of the Cs-I-Te-O-H-Ag-In-Cd-inert gas system. Calculations using RAFT show that under most accident conditions, the particle size spectrum is determined primarily by the competition between the homogeneous and heterogeneous nucleation mechanisms, rather than the agglomeration mechanism, and that direct vapor deposition on structural surfaces is an important mechanism for the scavenging of fission product vapors. 15 references, 8 figures.

Im, K.H.; Ahluwalia, R.K.; Chuang, C.F.

1985-01-01

72

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

73

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

SciTech Connect

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

Gauld, I.C.

2005-08-12

74

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

75

Measurement and characterization of fission products released from LWR fuel  

SciTech Connect

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

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

1984-01-01

76

Neutron capture and transmission measurements on fission product palladium-107  

Microsoft Academic Search

Neutron capture and transmission measurements were made on a sample of fission-product palladium. The level parameters were obtained for 34 ¹°⁷Pd resonances below 700 eV. The observed average level spacing was = (10.7 +- 1.5) eV, and the s-wave neutron strength function was determined to be 10⁴ S = (0.56\\/sup +0\\/.¹⁶\\/sub -0.12\\/). The capture width Gamma\\/sub ..gamma..\\/ for the

U. N. Singh; R. C. Block; Y. Nakagome

1978-01-01

77

Engineering Report on the Fission Gas Getter Concept  

SciTech Connect

In 2010, the Department of Energy (DOE) requested that a Brookhaven National Laboratory (BNL)-led team research the possibility of using a getter material to reduce the pressure in the plenum region of a light water reactor fuel rod. During the first two years of the project, several candidate materials were identified and tested using a variety of experimental techniques, most with xenon as a simulant for fission products. Earlier promising results for candidate getter materials were found to be incorrect, caused by poor experimental techniques. In May 2012, it had become clear that none of the initial materials had demonstrated the ability to adsorb xenon in the quantities and under the conditions needed. Moreover, the proposed corrective action plan could not meet the schedule needed by the project manager. BNL initiated an internal project review which examined three questions: 1. Which materials, based on accepted materials models, might be capable of absorbing xenon? 2. Which experimental techniques are capable of not only detecting if xenon has been absorbed but also determine by what mechanism and the resulting molecular structure? 3. Are the results from the previous techniques useable now and in the future? As part of the second question, the project review team evaluated the previous experimental technique to determine why incorrect results were reported in early 2012. This engineering report is a summary of the current status of the project review, description of newly recommended experiments and results from feasibility studies at the National Synchrotron Light Source (NSLS).

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

2012-11-01

78

On-line separation and identification of several short-lived fission products: Decay of 84Se, 91Kr, 97Y, 99Nb, 99Zr, 100, 101Nb and 101Zr  

Microsoft Academic Search

A device for nuclear spectroscopy of short-lived fission products was run at the focus of the gas-filled on-line mass separator at the FRJ-2 reactor. By the known y-lines of long-lived fission products, a mass calibration of the separator was carried out. Several unknown shortlived fission products could be identified by investigating some very intense gamma-lines. For these lines, the decay

J. Eidens; E. Roeckl; P. Armbruster

1970-01-01

79

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

80

Assessment of selected fission products in the Savannah River Site environment  

SciTech Connect

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

Carlton, W.H.; Denham, M.

1997-04-01

81

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

SciTech Connect

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

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

1982-08-01

82

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

83

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

84

Macroscopic Calculation al Model of Fission Gas Release from Water Reactor Fuels  

Microsoft Academic Search

Existing models for estimating fission gas release rate usually have fuel temperature as independent variable. Use of fuel temperature, however, often brings an excess ambiguity in the estimation because it is not a rigorously definable quantity as a function of heat generation rate and burnup. To derive a mathematical model that gives gas release rate explicitly as a function of

Masaaki UCHIDA

1993-01-01

85

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

86

Fission product scrubbing system for a nuclear reactor  

SciTech Connect

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

Leach, D.S.

1986-09-09

87

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

88

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

89

Corrosion of Fast-Reactor Claddings by Physical and Chemical Interaction with Fuel and Fission Products  

Microsoft Academic Search

Fuel-cladding chemical interaction in fast breeder reactor (FBR) fuel pins can cause both matrix and intergranular corrosion of the inner surface of the cladding. Matrix corrosion is uniform nonselective interaction with fuel and fission products, causing the cladding to thin. Intergranular corrosion occurs on grain boundaries, weakening both them and the grains. Interaction with fission products may be the cause

V. A. TZYKANOV; V. N. GOLOVANOV; V. K. SHAMARDIN; F. N. KRYUKOV; A. V. POVSTYANKO

90

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

Microsoft Academic Search

The fission products Cs-137 and Sr-90 are amongst the most common radionuclides occurring in ground contamination at the UK civil nuclear sites. Such contamination is often associated with alkaline liquids and the mobility of these fission products may be affected by these chemical conditions. Similar geochemical effects may also result from cementitious leachate associated with building foundations and the use

S. Kwong; J. Small

2007-01-01

91

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

SciTech Connect

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small - especially for {sup 99}Mo - we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for {sup 95}Zr, {sup 140}Ba, {sup 144}Ce), but are larger for {sup 99}Mo (4%-relative) and {sup 147}Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the {sup 147}Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the energy dependence over the fast neutron energy range from 0.2-2 MeV. Based on these trends, we present an evaluation of the FPY data at 0.5 and 2.0 MeV average incident neutron energies. This new set of ENDF/B-VII data will enable users to linearly interpolate between the pooled FPY data at {approx}0.5 MeV and our new data at 2 MeV to obtain FPYs at other energies.

Chadwick, M.B., E-mail: mbchadwick@lanl.go [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Kawano, T.; Barr, D.W.; Mac Innes, M.R.; Kahler, A.C.; Graves, T.; Selby, H.; Burns, C.J.; Inkret, W.C.; Keksis, A.L.; Lestone, J.P.; Sierk, A.J.; Talou, P. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2010-12-15

92

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

93

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

94

HEALTH HAZARDS FROM FISSION PRODUCTS AND FALLOUT. II. GAMMA RADIATION FROM NUCLEAR WEAPONS FALLOUT  

Microsoft Academic Search

Methods of estimating the gamma radiation from fallout and fission ; products are discussed. (Gamma spectra and 0.1 Mev energy intervals for 101 ; fissions were calculated for the following times after fission: 1. 2. 5. and 10 ; hours; years. Fifty energy groups of width 1.300 to 1.399 Mev were used, ; covering the range 0.0 to 5.0 Mev.

Bjornerstedt

1960-01-01

95

RAFT: a computer model for formation and transport of fission product aerosols in LWR primary system  

SciTech Connect

A computer model RAFT (Reactor Aerosol Formation and Transport) has been developed to predict the size distribution and composition of the particles (aerosols) formed from condensation of the fission-product and control rod material vapors released in LWR accidents. The condensation calculations in RAFT are driven by the equilibrium gas phase chemistry of Cs-I-Te-Ag-In-Cd-H-O-Ar system. The formation of the particles is considered to result as the gas becomes supersaturated with Ag, In, Cd, CsI, CsOH, Te/sub 2/ or TeO/sub 2/ vapor leading to the initiation of homogeneous nucleation. The rate of change in particle size spectrum as a result of homogeneous nucleation, heterogeneous nucleation, agglomeration, convection and deposition is described by the transient population balance equation.

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

1984-01-01

96

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

97

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

98

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

SciTech Connect

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

Selby, H.D., E-mail: hds@lanl.go [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2010-12-15

99

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

100

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

SciTech Connect

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

Beyer, Carl E.; Turnbull, Andrew J.

2010-01-29

101

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

102

Zirconium and fission product management in the ALSEP process  

SciTech Connect

Solvent extraction systems that combine neutral donor extractants and acidic extractants are being investigated to provide a single process solvent for separating Am and Cm from acidic high-level liquid waste, including their separation from the trivalent lanthanides. This approach of combining extractants is collectively referred to as the Actinide-Lanthanide Separation (ALSEP) process. Managing Zr and other fission products is one of the critical factors in developing the ALSEP process. In this work, a strategy has been developed in which Zr(IV) is extracted into the process solvent, then it is stripped from the solvent after the actinides have been selectively stripped. The ALSEP solvent contains a bifunctional neutral donor extractant that extracts the minor actinides and the trivalent lanthanides (Ln) from nitric acid media. In this work, two such extractants were considered: N,N,N',N'- tetraoctyl-diglycolamide (TODGA) and N,N,N',N'-tetra(2- ethylhexyl)diglycolamide (T2EHDGA). Molybdenum is strongly extracted into ALSEP solvents. Scrubbing the solvent with a citrate buffer before the actinide stripping step effectively removes Mo. Distribution ratios for Ru and Fe are low for extraction from HNO{sub 3}, so these components can easily be routed to the high-level waste raffinate. (authors)

Lumetta, G.J.; Carter, J.C.; Niver, C.M. [Pacific Northwest National Laboratory: P.O. Box 999, MSIN P7-25, Richland, WA 99352 (United States)

2013-07-01

103

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

104

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

E-print Network

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

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

2008-01-29

105

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

SciTech Connect

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

Karnaukhov, V. A. [Joint Institute for Nuclear Research (Russian Federation); Oeschler, H. [Darmstadt University of Technology, Institut fuer Kernphysik (Germany); Budzanowski, A. [H. Niewodniczanski Institute of Nuclear Physics (Poland); Avdeyev, S. P. [Joint Institute for Nuclear Research (Russian Federation); Botvina, A. S. [Institute for Nuclear Research (Russian Federation); Cherepanov, E. A. [Joint Institute for Nuclear Research (Russian Federation); Karcz, W. [H. Niewodniczanski Institute of Nuclear Physics (Poland); Kirakosyan, V. V.; Rukoyatkin, P. A. [Joint Institute for Nuclear Research (Russian Federation); Skwirczynska, I. [H. Niewodniczanski Institute of Nuclear Physics (Poland); Norbeck, E. [University of Iowa (United States)

2008-12-15

106

Fission product behavior in the Peach Bottom HTGR  

Microsoft Academic Search

The Peach Bottom high temperature, gas-cooled nuclear reactor, a 115-MW(t) power station, was placed in commercial operation on June 1, 1967. Operation with the initial core ended in October 1969 after 452 equivalent full power days (EFPD) had elapsed. Power production resumed on July 14, 1970 with the second core, and continued through October 31, 1974. During this period, Core

A. P. Malinauskas; F. F. Dyer; L. L. Fairchild; W. J. Martin; R. P. Wichner

1977-01-01

107

Thermodynamic analysis of chemical states of fission products in uranium-zirconium hydride fuel  

NASA Astrophysics Data System (ADS)

The chemical state of fission products (FPs) in U+ZrH 1.60 fuel was studied from the thermodynamic point of view. Twenty most abundant FP elements were taken into account in the system of U-Zr-H-O-FP in which oxygen is treated as an impurity. The Thermo-Calc computer code was used to calculate the equilibrium state of the multi-phase and multi-component system. This calculation shows that yttrium, the alkaline earth metals (Ba, Sr) and most of the lanthanides prefer to form corresponding binary hydrides. Oxygen impurities in the system are likely to form a mixture of Y 2O 3, Pr 2O 3, Sm 2O 3 as well as Ce 2O 3 depending on their fission yields. With increasing of burn-up, only a slight decrease of the hydrogen potential in the fuel pin can be expected because of the very little consumption of hydrogen in the hydrogenation process of the FP. In the gas phase, H 2, Cs, Rb, CsRb as well as CsI are the main vapor species apart from the noble gases Xe and Kr. Solid swelling of the fuel due to formation of condensed phases was calculated as a function of burn-up.

Huang, Jintao; Tsuchiya, Bun; Konashi, Kenji; Yamawaki, Michio

2001-04-01

108

Gamma-Ray Spectra of Fission Products observed with Li-Drifted Germanium Detectors, (II)  

Microsoft Academic Search

The ?-ray spectra of short-lived fission products from thermal neutron irradiation of highly enriched U were observed with an encapsulated Li-drifted Ge ?-ray spectrometer. The spectra at various periods10 min, 30 min, 1,2,5,10 and 20 hrafter irradiation were measured up to about 1 meV. The relative activities of fission products at various periods after irradiation (10 min20 hr) were calculated

Noboru ?I; Isami TANABE; Yasuyoshi MATSUSHIMA

1967-01-01

109

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

110

A proton-driven, intense, subcritical, fission neutron source for radioisotope production  

Microsoft Academic Search

99mTc, the most frequently used radioisotope in nuclear medicine, is distributed as 99Mo&squflg;99mTc generators. 99Mo is a fission product of 235U. To replace the aging nuclear reactors used today for this production, we propose to use a spallation neutron source, with neutron multiplication by fission. A 150 MeV, H? cyclotron can produce a 225 kW proton beam with 50% total

Yves Jongen; Yves

1995-01-01

111

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

112

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

113

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

114

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

SciTech Connect

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

Deforest, D.L.

1991-12-01

115

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

116

SEARCH FOR HIGH ENERGY ?-DECAY FROM THE SPONTANEOUS FISSION PRODUCTS OF 252CF  

Microsoft Academic Search

We searched for high energy ?-particles emitted from spontaneous fission products of 252 Cf by the measurement of the high energy part of the ?-spectrum with a very low background ?E1-?E2-Et -scintillation spectrometer. It is found that ?-emissions in the energy region higher than 20 MeV exist with the relative probability of Y? = (4.50.6 stat.1.2 syst.) x10 -9 ?\\/fission.

Yu. V. Ryabov

117

Fission product beta and gamma energy release. Quarterly progress report for October--December 1975  

Microsoft Academic Search

Preliminary experimental information for beta-ray energy release from fission product decay following thermal-neutron fission of U has been obtained for cooling times between 3 and 14,400 secs. The data were obtained as pulse-height spectra for beta energies between 0.25 and 8 MeV using a two-crystal scintillation spectrometer, and were unfolded to give beta-ray energy spectra of moderate resolution. Two irradiation

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

1976-01-01

118

Fission product beta and gamma energy release. Quarterly progress report, July--September 1975  

Microsoft Academic Search

Preliminary data for gamma-ray energy release from fission product decay ; following thermal-neutron fission of U has been obtained and reduced for ; cooling times between 3 and 14,400 sec. The data were obtained as pulse-height ; spectra for photon energies between 0.05 and 8 MeV using a NaI spectrometer and ; were unfolded to give photon-energy spectra of moderate

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

1975-01-01

119

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

120

Continuous fission-product monitor system at Oyster Creek. Final report  

SciTech Connect

A continuous on-line fission product monitor has been installed at the Oyster Creek Nuclear Generating Station, Forked River, New Jersey. The on-line monitor is a minicomputer-controlled high-resolution gamma-ray spectrometer system. An intrinsic Ge detector scans a collimated sample line of coolant from one of the plant's recirculation loops. The minicomputer is a Nuclear Data 6620 system. Data were accumulated for the period from April 1979 through January 1980, the end of cycle 8 for the Oyster Creek plant. Accumulated spectra, an average of three a day, were stored on magnetic disk and subsequently analyzed for fisson products, Because of difficulties in measuring absolute detector efficiency, quantitative fission product concentrations in the coolant could not be determined. Data for iodine fission products are reported as a function of time. The data indicate the existence of fuel defects in the Oyster Creek core during cycle 8.

Collins, L.L.; Chulick, E.T.

1980-10-01

121

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

122

Shale gas production: potential versus actual greenhouse gas emissions*  

E-print Network

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

123

On the Radiochemical Separations of the Beta-emitting Fission Products  

NASA Astrophysics Data System (ADS)

This research aims at developing fast and effective radiochemical procedures for separation of the beta-emitting fission products that are difficult to analyze by gamma-spectrometry. Post-detonation analysis, as one of the major tasks of nuclear forensics, can provide crucial information for identification of the explosion levels, fuel sources, and industrial processes of a nuclear device. However, a dozen of radionuclides with high fission yields such as Zr-93, Tc-99, Sr-90 are either pure beta-emitters or only emitting gamma-rays that are difficult to analyze. Although the analysis of these radionuclides was thoroughly studied, samples from unknown nuclear detonations can be complicated by the number of fission products, radioactivity levels, sample matrices, and time limits for analysis. The challenge facing the forensic analysis should not be underestimated. A sequential separation procedure is designed to analyze the major beta-emitting fission products. Radiochemical techniques such as solvent extraction, precipitation, and column chromatography are utilized. The procedure will be tested and improved by experiments. The final procedure should be capable of analyzing the fission products under various sample conditions effectively and rapidly.

Chang, Zheng; Sudowe, Ralf

2013-04-01

124

Performance of a zeolite column system in removing fission products from molten salt  

SciTech Connect

Spent nuclear fuel is dissolved in molten chloride salt and treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. Non-TRU fission products in the molten salt are removed by ion-exchange in a continuous process using a zeolite column system. The salt-loaded zeolite is subsequently mixed with glass and consolidated. The zeolite was effective at removing fission products from the molten salt. Cesium was removed relatively rapidly; alkaline earths and rare earths were more strongly held by the zeolite, but the rate of exchange was much slower. Four parameters were varied during these tests: type of zeolite used, temperature, flow rate, and composition of the salt. The effects of these parameters on the effluent composition and on the distribution of fission products along the length of the column are presented and discussed. Results are used to test a computer model of the system.

Babcock, B.; Pereira, C.; Hutter, J. [Argonne National Lab., IL (United States)

1996-10-01

125

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

126

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

127

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

128

Delayed neutron noise characteristics of an in-pile fission product loop  

SciTech Connect

This paper reports on delayed neutron noise measurements carried out in an in-pile sodium loop, the Fission Product Loop 2 (FPL-2), installed on the Toshiba Training Reactor I. To clarify the characteristics and origin of delayed neutron count rate noise, a noise propagation mechanism was identified using a multivariate autoregressive model. The results show that a simulated fuel failure in the FPL-2, with recoil as the principal fission product release phenomena, produces a white spectrum of delayed neutron count rate noise. It was also found that the loop temperature fluctuation strongly affects the delayed neutron count rate noise at temperatures below 300{degrees} C, through the deposition of fission products on the surface of structures.

Tamaoki, T. (Toshiba Corp., Nuclear Engineering Lab., 8 Shin-Sugita, Isogo-ku, Yokohama 235 (JP)); Sakai, T.; Endo, H. (Toshiba Corp., Isogo Engineering Center, 8 Shin-Sugita, Isogo-ku, Yokohama 235 (JP)); Haga, K. (Power Reactor and Nuclear Fuel Development Corp., O-aral Engineering Center, 4002 Narita, O-araimachi, Ibaraki-ken 311-13 (JP)); Takahashi, R. (Tokyo Inst. of Technology, Dept. of Mechanical Engineering, Ookayama, Meguro-ku, Tokyo 152 (JP))

1992-07-01

129

Diffusion of fission products and radiation damage in SiC  

NASA Astrophysics Data System (ADS)

A major problem with most of the present nuclear reactors is their safety in terms of the release of radioactivity into the environment during accidents. In some of the future nuclear reactor designs, i.e. Generation IV reactors, the fuel is in the form of coated spherical particles, i.e. TRISO (acronym for triple coated isotropic) particles. The main function of these coating layers is to act as diffusion barriers for radioactive fission products, thereby keeping these fission products within the fuel particles, even under accident conditions. The most important coating layer is composed of polycrystalline 3C-SiC. This paper reviews the diffusion of the important fission products (silver, caesium, iodine and strontium) in SiC. Because radiation damage can induce and enhance diffusion, the paper also briefly reviews damage created by energetic neutrons and ions at elevated temperatures, i.e. the temperatures at which the modern reactors will operate, and the annealing of the damage. The interaction between SiC and some fission products (such as Pd and I) is also briefly discussed. As shown, one of the key advantages of SiC is its radiation hardness at elevated temperatures, i.e. SiC is not amorphized by neutrons or bombardment at substrate temperatures above 350 C. Based on the diffusion coefficients of the fission products considered, the review shows that at the normal operating temperatures of these new reactors (i.e. less than 950 C) the SiC coating layer is a good diffusion barrier for these fission products. However, at higher temperatures the design of the coated particles needs to be adapted, possibly by adding a thin layer of ZrC.

Malherbe, Johan B.

2013-11-01

130

Experimental measurement of distribution coefficients of Pd and Sr from a synthetic mixed fission product solution  

SciTech Connect

The extraction of palladium and strontium from a synthetic non-radioactive fission product solution in 3 M nitric acid was studied. The experimental distribution coefficients (org/aq) of all fifteen fission products were determined using liquid ion exchange reagents and a new crown compound in an organic phase of tri-n-butyl phosphate (TBP) and kerosene. The most effective system tested for the extraction of Pd from the synthetic fission product solution in 3 M nitric acid was determined to be a 0.3 M solution of the tertiary amine, Alamine 336, in an organic phase of TBP and kerosene having a TBP/kerosene volumetric ratio of 0.667. Contact of this organic solution with an equal volume of synthetic fission product solution yielded a palladium distribution coefficient (org/aq) of 1.95. The new crown compound was found to selectively extract Sr and Ba from the synthetic fission product solution. The new crown, 4,4'(5)(1-hydroxyheptyl)cyclohexo-18-crown-6 (Crown XVI), was formed by hydrogenating the benzene rings attached to the crown ring of the parent compound, bis(4,4'(5')-(1-hydroxyheptyl)benzo)-18-crown-6 (Crown IX). The Crown IX had been shown in previous work to be selective for Cs. Distribution coefficients (org/aq) of 2.55 and 2.85 for Sr and Ba, respectively, were obtained when the synthetic fission product solution was contacted with an equal volume of organic solution containing 0.02 M Crown XVI in 5 vol % didodecylnaphthalene sulfonic acid (DNS)/27 vol % TBP/68 vol % kerosene.

Shuler, R.G.

1982-01-01

131

Distribution of fission products in the homogeneous liquid-liquid extraction of uranium  

SciTech Connect

Separation of uranium from fission products by homogeneous liquid-liquid extraction of uranium from one molar nitric acid solution with addition of ferric nitrate as salting-out reagent, into propylene carbonate has been performed. Uranium(VI) was quantitatively extracted into propylene carbonate from an aqueous medium of 0.5 g/l Fe(NO{sub 3}){sub 3} 9H{sub 2}O and 1 M HNO{sub 3} at 99 C, then quantitatively stripped from the organic phase with 0.1 M sodium carbonate at pH 9. Final separation of uranium(VI) was obtained by extracting uranium(VI) into 0.1 M dibenzoyl methane in propylene carbonate using the homogeneous technique at pH 7. Precipitation of ferric hydroxide affords efficient decontamination from significant fission products. The representative fission product elements, molybdenum, strontium, ruthenium, zirconium, and cerium, remained in the aqueous solution after after extracting uranium(VI) into propylene carbonate to an extent grater than 97%; i.e., less than three percent of the respective elements were found in the carbonate stripping solution. After the final separation step, the extraction of uranyl ion into propylene carbonate containing dibenzoly methane, these fission product elements were not longer detectable. Ten percent of the original concentration of iodide was found in the carbonate stripping solution. However, it was removed in the final separation step. This uranium extraction method can be applied as a practical method for separating uranium from fission products to recover the uranium from spent fuel elements. The capacity of ferric hydroxide for adsorption of fission products and the ability to convert to the somewhat refractory ferric oxide also promises convenience for long term storage.

Xu, J.

1988-01-01

132

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. PMID:3202634

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

1988-01-01

133

ACTIVE MEDIA: Influence of the gas stream velocity on the output power of gas-flow lasers excited by fission fragments of uranium nuclei  

Microsoft Academic Search

A model is proposed for gas-flow lasers excited by fission fragments of uranium nuclei. This model describes the dependence of the active lasing volume on the density of the gas mixture, on the velocity of the gas stream, and on the pump power. Calculations based on this model are in satisfactory agreement with the experimental time characteristics of the output

A. N. Sizov; V. V. Porkhaev

1996-01-01

134

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

SciTech Connect

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

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

1984-01-01

135

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

SciTech Connect

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

Wichner, R.P.

1983-01-01

136

Experimental Determination of the Antineutrino Spectrum of the Fission Products of $^{238}$U  

E-print Network

An experiment was performed at the scientific neutron source FRM II in Garching to determine the cumulative antineutrino spectrum of the fission products of $^{238}$U. This was achieved by irradiating target foils of natural uranium with a thermal and a fast neutron beam and recording the emitted $\\beta$-spectra with a gamma-suppressing electron-telescope. The obtained $\\beta$-spectrum of the fission products of $^{235}$U was normalized to the data of the magnetic spectrometer BILL of $^{235}$U. This method strongly reduces systematic errors in the $^{238}$U measurement. The $\\beta$-spectrum of $^{238}$U was converted into the corresponding antineutrino spectrum. The final $\\bar\

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

2013-01-01

137

Experimental Determination of the Antineutrino Spectrum of the Fission Products of $^{238}$U  

E-print Network

An experiment was performed at the scientific neutron source FRM II in Garching to determine the cumulative antineutrino spectrum of the fission products of $^{238}$U. This was achieved by irradiating target foils of natural uranium with a thermal and a fast neutron beam and recording the emitted $\\beta$-spectra with a gamma-suppressing electron-telescope. The obtained $\\beta$-spectrum of the fission products of $^{235}$U was normalized to the data of the magnetic spectrometer BILL of $^{235}$U. This method strongly reduces systematic errors in the $^{238}$U measurement. The $\\beta$-spectrum of $^{238}$U was converted into the corresponding antineutrino spectrum. The final $\\bar\

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

2013-12-19

138

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

139

Method of Fission Product Beta Spectra Measurements for Predicting Reactor Anti-neutrino Emission  

E-print Network

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

D. M. Asner; K. Burns; L. W. Campbell; B. Greenfield; M. S. Kos; J. L. Orrell; M. Schram; B. VanDevender; 1 L. S. Wood; D. W. Wootan

2014-03-01

140

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

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

141

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

Microsoft Academic Search

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

J. P. Ackerman; T. R. Johnson

1993-01-01

142

Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs  

SciTech Connect

Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design.

Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

1980-11-01

143

DETERMINATION OF NEPTUNIUM IN URANIUM--FISSION PRODUCT MIXTURES. INITIAL EXTRACTION WITH METHYL ISOBUTYL KETONE  

Microsoft Academic Search

A method for the separation and determination of neptunium in uranium--; fission product mixtures uses a twocycle extraction system. Neptunium is ; oxidized to 6+ with permanganate and quantitatively extracted as a nitrate ; complex into methyl isobutyl ketone from an acid-deficient aluminum nitrate ; salting solution containing tetrapropylammonium nitrate. Neptunium is stripped ; from the ketone phase and simultaneously

W. J. Maeck; G. L. Booman; M. C. Elliott; J. E. Rein

1960-01-01

144

Modeling of a zeolite column for the removal of fission products from molten salt  

SciTech Connect

During electrorefining of spent nuclear fuels, fission products, and actinides accumulate in the LiCl-KCl electrolyte salt. From the standpoint of high-level waste minimization, it is advantageous to remove these from the salt rather than discarding the salt after these have built up to certain concentrations. Laboratory experiments have shown that zeolite A has the desirable properties for selective removal of fission products by the ion-exchange process. An analytical model has been developed to assess the performance of the ion-exchange column and to scale up the data to engineering size equipment. The model employs empirically determined exchange factors to represent the equilibrium chemistry of the exchange process. The exchange kinetics is modeled by the Nernst-Planck equation applied to multicomponent diffusion in the zeolite pellet. The self-diffusion coefficients, derived from batch kinetic data, are substantially different for monovalent, divalent, and trivalent fission products. The method of characteristics is used to solve the transport equations for the salt phase. The computed breakthrough of the fission products through the experimental zeolite column is in partial but not complete agreement with the measured concentration profiles.

Ahluwalia, R.K.; Geyer, H.K.; Pereira, C.; Ackerman, J.P. [Argonne National Lab., IL (United States)] [Argonne National Lab., IL (United States)

1998-01-01

145

Neutron and gamma spectra in delayed neutron decay of some fission products  

Microsoft Academic Search

From spring meeting of the Deutsche Physikalische Gesellschaft. Part A, ; nuclear physics; Heidelberg, Germany (19 Feb By use of fast radiochemical ; separation methods, (separation time: 24 sec), several short-lived isotopes of ; As, Se, Br, Te, and I were isolated from ²³⁵U fission products and ; identified as emitters of delayed neutrons. With the new high pressure, high

H. Folger; H. Franz; W. Grimm; G. Hermann; J. V. Kratz; K. L. Kratz; F. Nuh; S. G. Prussin

1973-01-01

146

CHEMICAL EFFECTS IN FISSION PRODUCT RECOIL. III. THE DECOMPOSTION OF POTASSIUM NITRATE  

Microsoft Academic Search

The effect of the recoil of fission products from uranium foils inio an ; adjacent layer of potassium nitrate was investigated. Under these conditions ; about 10.5 nitrite ions are formed per 100 ev of recoil energy absorbed. This is ; considerably greater than the values observed for other forms of radiation and ; the reasons for this are discussed.

D. Hall; G. N. Walton

1958-01-01

147

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

E-print Network

used, #12;- 4 - Plant Species Forty-four different plant species were tested. In table 3 the com- monNovember, 1967 Riso Report No. 170 Investigations on the Plant Uptake of Fission Products from Contaminated Soils. I. Influence of Plant Species and Soil Types on the Uptake of Radioactive Strontium

148

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

Microsoft Academic Search

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

Gohar

2000-01-01

149

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...

150

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

151

Method of Fission Product Beta Spectra Measurements for Predicting Reactor Anti-neutrino Emission  

E-print Network

The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron rich fission products that subsequently beta decay and emit electron anti-neutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to current precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent re-considerations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable i...

Asner, D M; Campbell, L W; Greenfield, B; Kos, M S; Orrell, J L; Schram, M; VanDevender, B; Wood, 1 L S; Wootan, D W

2014-01-01

152

Shale gas production: potential versus actual greenhouse gas emissions  

E-print Network

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

OSullivan, Francis Martin

153

Thermal release of volatile fission products from irradiated nuclear fuel  

Microsoft Academic Search

An effective procedure for removing H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO fuel. The release characteristics of H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released

L. A. Bray; L. L. Burger; L. G. Morgan; D. L. Baldwin

2011-01-01

154

Some aspects of the separation of nuclear fission products by liquid-liquid extraction  

Microsoft Academic Search

The technique for and methods of separation of products of nuclear fission play a major role in many stages of the nuclear\\u000a fuel cycle. The extraction of these products from effluent solution after the processing of the burnt-up nuclear fuel is receiving\\u000a considerable attention. Trivalent lanthanoides are usualy extracted together with Am(III) and their mutual separation is rather\\u000a difficult.14 The

V. Jedinkov

1983-01-01

155

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor1  

Microsoft Academic Search

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% 235U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo\\/99mTc generators

A. Mushtaq; Massod Iqbal; Ishtiaq Hussain Bokhari; Tariq Mahmood; Tayyab Mahmood; Zahoor Ahmad; Qamar Zaman

2008-01-01

156

Isomers in Fission Fragments  

Microsoft Academic Search

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

W. Urban; H. Faust; M. Jentschel; U. Koester; J. Krempel; Th. Materna; P. Mutti; T. Soldner; J. Genevey; J. A. Pinston; G. Simpson; T. RzaCa-Urban; A. Zlomaniec; M. Lukasiewicz; K. Sieja; F. Nowacki; O. Dorvaux; B. J. P. Gall; B. Roux; J. A. Dare; J. L. Durell; A. G. Smith; B. J. Varley; I. Tsekhanovich; J. Jolie; A. Linnemann; A. Scherillo; R. Orlandi; J. F. Smith; I. Ahmad

2009-01-01

157

Data summary report for fission product release test VI6  

Microsoft Academic Search

Test VI-6 was the sixth test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the BR3 reactor in Belgium. The fuel had experienced a burnup of 42 MWd\\/kg, with inert gas release during irradiation of 2%. The fuel specimen was heated in an induction furnace at 2300

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

1994-01-01

158

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

159

Production of Mass-Separated Fission Fragment Beams at ALTO  

SciTech Connect

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

Lebois, M.; Cheikh Mhamed, M.; Curaudeau, J. M.; Ducourtieux, M.; Essabaa, S.; Franchoo, S.; Gales, S.; Guillemaud-Mueller, D.; Ibrahim, F.; Lau, C.; Lesrel, J.; Mueller, A.; Raynaud, M.; Roussiere, B.; Said, A.; Verney, D.; Vogel, C. [Institut de Physique Nucleaire, CNRS-IN2P3/Univ. Paris Sud-XI, F-91406 Orsay Cedex (France)

2007-05-22

160

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

SciTech Connect

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

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

1982-01-01

161

Measurement of airborne fission products in Chapel Hill, NC, USA from the Fukushima I reactor accident  

E-print Network

We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days following the March 11, 2011, accident at the Fukushima I 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 I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^2 and 0.42 +/- 0.07 mBq/m^2 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

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

2011-01-01

162

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

163

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

SciTech Connect

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

Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

2012-07-01

164

Modeling of a zeolite column for the removal of fission products from molten salt  

Microsoft Academic Search

During electrorefining of spent nuclear fuels, fission products, and actinides accumulate in the LiCl-KCl electrolyte salt. From the standpoint of high-level waste minimization, it is advantageous to remove these from the salt rather than discarding the salt after these have built up to certain concentrations. Laboratory experiments have shown that zeolite A has the desirable properties for selective removal of

R. K. Ahluwalia; H. K. Geyer; C. Pereira; J. P. Ackerman

1998-01-01

165

STUDIES OF SHORT-LIVED FISSION PRODUCTS AND THEIR IMPORTANCE TO REACTOR TECHNOLOGY  

Microsoft Academic Search

A systematic study of the decay schemes of some of the important alkali ; metal isotopes was made. Information is available on 17.8-minute Rb⁸⁸, ; 14.9-minute Rb⁸⁹, and 2.6-minute Rb⁹°. The decay characteristics of ; these nuclides show the general features exhibited by all of the short-lived ; fission products studied so far, namely, the short half-lives are related to

G. D. OKelley; E. Eichler; N. R. Johnson

1958-01-01

166

Decay energies of gaseous fission products and their daughters for A = 138 to 142  

Microsoft Academic Search

The BETA -decay energies For several mass-separated Xe fission products ; and their daughters have been measured at the TRISTAN on-line separator facility ; at the Ames Laboratory Research Reactor. A well-type plastic scintillator was ; used in coincidence with a Ge(Li) gamma detector to detennine BETA -group end-; point energies and deduce Q values. The follow:ng BETA -decay energies

J. P. Adams; G. H. Carlson; M. A. Lee; W. L. Jr. Talbert; F. K. Wohn; J. R. Clifford; J. R. McConnell

1973-01-01

167

Decay Energies of Gaseous Fission Products and their Daughters for A=88 to 93  

Microsoft Academic Search

A systematic study of beta-decay energies has been made for mass-separated activities of Kr gaseous fission products and their daughters at the TRISTAN on-line separator facility at the Ames Laboratory research reactor. A well-type plastic scintillator was used in coincidence with a Ge(Li) gamma detector to determine beta-group end-point energies and deduce Q values. The following beta-decay energies have been

J. R. Clifford; W. L. Talbert; F. K. Wohn; J. P. Adams; J. R. McConnell

1973-01-01

168

Decay Energies of Gaseous Fission Products and their Daughters for A=138 to 142  

Microsoft Academic Search

The beta-decay energies for several mass-separated Xe fission products and their daughters have been measured at the TRISTAN on-line separator facility at the Ames Laboratory research reactor. A well-type plastic scintillator was used in coincidence with a Ge(Li) gamma detector to determine beta-group end-point energies and deduce Q values. The following beta-decay energies have been determined: 138Xe, 2.83 +\\/- 0.08

J. P. Adams; G. H. Carlson; M. A. Lee; W. L. Talbert; F. K. Wohn; J. R. Clifford; J. R. McConnell

1973-01-01

169

Neutron Cross Section Evaluation of Fission Products in the Fast Energy Region  

Microsoft Academic Search

Neutron cross sections of 19 priority fission products were evaluated from the unresolved resonance energy region up to 20 MeV. The present work complements the evaluation below the fast neutron energy region within the framework of the KAERI-BNL collaboration (1). The project was motivated by the need to improve ENDF\\/B-VI evaluation for a number of applications, including the burn-up credit

YongDeok Lee; Jonghwa Chang; Pavel Oblozinsky

2000-01-01

170

RADIATION DAMAGE TO AMINE EXTRACTANTS. I. EFFECT ON THE EXTRACTION BEHAVIOR OF URANIUM AND FISSION PRODUCTS  

Microsoft Academic Search

Radiation damage to 5 vol% alkyl amine-kerosene modified with lauryl ; alcohol was studied through the behavior changes of uranium and some fission ; products in extraction, scrubbing, and stripping in the nitrate system. Tested ; were three tertiary and four secondary amines, i.e., trioctyl amine, N-benzyl ; dilauryl amine, N-cyclohexyl dilauryl amine, N-cyclohexyl lauryl amine, dilauryl ; amine, Amberlite

T. Ishihara; T. Tsujino; Y. Komaki

1962-01-01

171

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

SciTech Connect

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 the fuel; second, the measured vaporization rate of the various species makes it possible to check the thermodynamic models of the system fuel + fission products, used to predict the chemical reactions occurring during reactor accidents and, hence, the state of the radiotoxical nuclides released. Here, irradiated light water reactor fuel from the BR3 reactor was thermally annealed up to 2,500 K in a Knudsen cell, and the effusing vapors were measured by mass spectrometry. The experiments provide data on the stoichiometry evolution of the fuel during release as well as a reliable method to evaluate the diffusion coefficients of volatile and less-volatile fission products. The analysis of the data starts from diffusion of xenon, which clearly shows three typical release stages respectively controlled by radiation damage annealing, self-diffusion, and matrix vaporization. The experimental measurements are also in agreement with the predictions of intragranular trapping models. Barium and cesium showed faster release than xenon, the former being likely to diffuse atomically to the grain boundaries where no evidence of formation of stable zirconates was found. These results were compared with those obtained by a burnup-simulated fuel, where barium was initially present in a perovskite phase, producing essentially different release patterns.

Capone, F.; Hiernaut, J.P.; Martellenghi, M.; Ronchi, C. [European Commission, Karlsruhe (Germany). European Inst. for Transuranium Elements

1996-11-01

172

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

173

Electrochemical separation of actinides and fission products in molten salt electrolyte  

Microsoft Academic Search

Molten salt electrochemical separation may be applied to accelerator-based conversion (ABC) and transmutation systems by dissolving the fluoride transport salt in LiCl-KCl eutectic solvent. The resulting fluoride-chloride mixture will contain small concentrations of fission product rare earths (La, Nd, Gd, Pr, Ce, Eu, Sm, and Y) and actinides (U, Np, Pu, Am, and Cm). The Gibbs free energies of formation

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

1995-01-01

174

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter  

Microsoft Academic Search

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of\\u000a PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and\\u000a a LLFP transmutation zone (TZ) containing the 99TC and 129I and 135Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic

Hseyin Yap?c?; Gl?ah z???k

2008-01-01

175

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

Code of Federal Regulations, 2010 CFR

...marketable production of natural gas on an as sold basis. Production...include dry, residue, and wet gas, depending on whether liquids have been extracted before the registrant transfers title. Flared gas, injected gas, and gas...

2010-04-01

176

Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization  

SciTech Connect

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

Joshua R. Versey

2013-08-01

177

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

SciTech Connect

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

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

1988-01-01

178

A Strategic Recovery of Rare-Metal Fission Products in Spent Nuclear Fuel  

SciTech Connect

Based on state-of-the-art separation chemistry, extended recycling of rare-metal fission products (RMFPs) from fast breeder reactors is examined as a new strategy for spent fuel reprocessing. Fission product fractionation is in accordance with the modern trend toward zero emission of toxic materials; salt-free separation utilizing in situ electrochemistry will suit the current direction of research and development in the back end of the nuclear fuel cycle. A catalytic electrolytic extraction and dissolution method, which would avoid secondary waste arising, was proposed to separate the target, the radioactive but potentially strategic elements Pd, Ru, Rh, Tc, Te, and Se, dissolved in high-level liquid waste (HLLW). It was confirmed that RMFPs could be recovered essentially from simulated HLLW with the conceptual scheme, although further studies for the optimization were required to obtain higher recovery ratios of RMFPs. Elemental separation not only offers alternative material resources to meet expanding demands for catalysts in fuel cell/hydrogen energy systems but is also the first step for transmutation or other selective strategies for waste management of long-lived fission products.

Sano, Yuichi; Shinoda, Yoshihiko; Ozawa, Masaki [Japan Nuclear Fuel Cycle Development Institute (Japan)

2004-12-15

179

RADIATIONS FROM SHORT-LIVED RARE GAS FISSION PRODUCTS  

Microsoft Academic Search

The energies of the primary gamma rays emitted in the decay of 3.2 min ; Kr⁸⁹, 33 sec Kr⁹°, 1.2 min Rb\\/sup 91m\\/, 41 sec Xe\\/sup 139 and 66 sec ; Cs¹⁴° were determined and relative gamma -ray intensities measured. ; Photon per disintegration values were estimated for Kr⁸⁹, Kr⁹°, Xe\\/; sup 139\\/, and for 9.5 min Cs¹³⁹. BETA -ray

M. A. Wahlgren; W. W. Meinke

1962-01-01

180

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

181

Fission-product yields for thermal-neutron fission of ²⁴³Cm determined from measurements with a high-resolution low-energy germanium gamma-ray detector  

Microsoft Academic Search

Cumulative fission-product yields have been determined for 13 gamma rays emitted during the decay of 12 fission products created by thermal-neutron fission of ²⁴³Cm. A high-resolution low-energy germanium detector was used to measure the pulse-height spectra of gamma rays emitted from a 77-nanogram sample of ²⁴³Cm after the sample had been irradiated by thermal neutrons. Analysis of the data resulted

1984-01-01

182

Thermochemical Prediction of Chemical Form Distributions of Fission Products in LWR Oxide Fuels Irradiated to High Burnup  

Microsoft Academic Search

Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial

Kouki MORIYAMA; Hirotaka FURUYA

1997-01-01

183

Behavior of U-Zr Alloy Containing Simulated Fission Products during Anodic Dissolution in Molten Chloride Electrolyte  

Microsoft Academic Search

To investigate the distribution of fission product elements in the electrorefining of spent metallic fuel, electrorefining tests were carried out using U-Zr alloy containing simulated fission products. An intermediate region was found between undissolved alloy and Zr-rich regions formed by the anodic dissolution of uranium. The composition of this region determined by wavelength-dispersive X-ray spectroscopy (WDS) analysis corresponded to that

Masatoshi IIZUKA; Takashi OMORI; Takeshi TSUKADA

2010-01-01

184

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

Microsoft Academic Search

The discrepancies between experimental findings and the Reactor Safety Study predictions, as well as the rapidly developing data base enabling phenomenological modeling of core-concrete interactions and ex-vessel fission product release, have prompted the development of several new computer models of core-concrete interactions and fission product release during severe accidents. Two such models are the CORCON-MOD2 model of core-concrete interactions and

G. A. Greene; Y. Sanborn

1986-01-01

185

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

186

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

187

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

NASA Astrophysics Data System (ADS)

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

Juhasz, Albert J.

188

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

Code of Federal Regulations, 2013 CFR

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

2013-04-01

189

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

Code of Federal Regulations, 2011 CFR

... 2011-04-01 false (Item 1204) Oil and gas production, production prices and...S-K Disclosure by Registrants Engaged in Oil and Gas Producing Activities 229.1204 (Item 1204) Oil and gas production, production prices...

2011-04-01

190

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

Code of Federal Regulations, 2012 CFR

... 2012-04-01 false (Item 1204) Oil and gas production, production prices and...S-K Disclosure by Registrants Engaged in Oil and Gas Producing Activities 229.1204 (Item 1204) Oil and gas production, production prices...

2012-04-01

191

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

Microsoft Academic Search

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

B. Nagy; M. J. Rigali

1993-01-01

192

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

193

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

194

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

195

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

196

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

Microsoft Academic Search

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

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

1978-01-01

197

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

198

Mass-Yield Distribution of the Fission Products in Fallout from the 14 May 1965 Nuclear Explosion  

Microsoft Academic Search

Twenty single particles separated from a 20-liter sample of rain collected in Osaka, Japan, shortly after the 14 May 1965 test explosion of the Chinese nuclear device, were analyzed radiochemically. The abundance pattern of the fission products in these particles resembled the shape of the mass-yield curve for the thermal neutron-induced fission of uranium-235, except for the facts that cesium-h137

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

1966-01-01

199

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

200

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

SciTech Connect

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

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

2005-09-01

201

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

NASA Astrophysics Data System (ADS)

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

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

2008-10-01

202

Nuclear Fission  

Microsoft Academic Search

The probability of nuclear fission is reviewed relative to spontaneous ; fission half lives, penetration of the fission barrier, fission with siow ; neutrons, fission at inoderate and high excitation energies. fission cross ; sections near the threshold, and the fission of elements lighter than thorium. ; The energy available for fission and the kinetic and excitation energies of ;

I. Halpern

1959-01-01

203

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

204

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

NASA Astrophysics Data System (ADS)

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

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

2006-05-01

205

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

NASA Astrophysics Data System (ADS)

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

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

2009-06-01

206

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

SciTech Connect

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

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

2013-02-01

207

High-accuracy mass spectrometry of fission products with Penning traps  

NASA Astrophysics Data System (ADS)

Mass measurements of fission products based on Penning-trap technique are reviewed in this article. More than 300 fission products have been measured with JYFLTRAP, ISOLTRAP, CPT, LEBIT and TITAN Penning traps with a typical precision of ?m/m ? 10-7 - 10-8. In general, the results agree well with each other. The new data provide a valuable source of information and a challenge for the future development of theoretical mass models as well as for obtaining a deeper insight into microscopic properties of atomic nuclei as measured, for example, via key mass differentials. Shape transitions around N = 60, subshell closure at N = 40 and shell closures at N = 50 and N = 82 have been investigated in the trend of the precisely measured two-neutron separation energies. The evolution of two-neutron and two-proton shell gaps has been studied and compared to theoretical models for Z = 50, N = 50 and N = 82. Proton-neutron pairing effect in separation energies and odd-even staggering of masses are shortly discussed. In addition to nuclear structure, many experiments have been motivated by nuclear astrophysics.

Kankainen, A.; yst, J.; Jokinen, A.

2012-09-01

208

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

209

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

E-print Network

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

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

2012-08-19

210

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

211

Generation of lumped fission product cross sections for high burnup, highly enriched uranium fuel  

SciTech Connect

The first set of reactor design calculations for the reactor design considered here was performed with a depletion methodology developed for converter reactor studies. These analyses showed that the ANS reactor would have a cycle length of 14 days when operated at a power level of 270 MW. Since both the cycle length and the discharge fuel burnup (209,000 MWD/MT) are very different from any of the reactors for which the depletion methodology was developed, a new study of the depletion process was initiated. Since the expected cycle length and fuel loading (18.1 kg /sup 235/U) were known, input for an ORIGEN calculation could be prepared. For the work described here, cross section updates for the actinides and major fission products were prepared with data from an ENDF/B-V-derived library. The NITAWL-S and XSDRNPM-S codes were used to perform this update. The XSDRNPM model was a one-dimensional, buckled, cylindrical representation of the reactor. Fission yield values were derived from ENDF/B-IV data as contained in the ORIGEN Pressurized Water Reactor Library. 9 refs., 2 figs.

Primm, R.T. III; Greene, N.M.

1988-01-01

212

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter  

NASA Astrophysics Data System (ADS)

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the 99TC and 129I and 135Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of 6Li in lithium (Li) and the theoretical density (TD) of Li2O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ? 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.

Yap?c?, Hseyin; z???k, Gl?ah

2008-12-01

213

Natural gas hydrates - issues for gas production and geomechanical stability  

E-print Network

Studies of Texas A&M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY Approved by: Co-Chairs of Committee, Stephen A Holditch George J Moridis Committee Members, William D McCain Maria Barrufet... that the hydraulic fracture gets plugged by the formation of secondary hydrates during gas production. I used the coupled fluid flow and geomechanical model TOUGH+Hydrate- FLAC3D to model geomechanical performance during gas production from hydrates...

Grover, Tarun

2008-10-10

214

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

215

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

NASA Astrophysics Data System (ADS)

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

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

2008-04-01

216

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

217

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

218

Application of ionic liquids in actinide and fission product separations: progress and prospects.  

SciTech Connect

Ionic liquids (ILs), particularly those that are liquid at room temperature, have attracted intense interest as alternatives to conventional organic solvents in a host of synthetic, catalytic, and electrochemical applications. Recently, growing attention has been devoted to their use in separations, typically as replacements for the organic diluents employed in traditional liquid-liquid extraction or membrane-based separations of organic solutes or metal ions. Although studies of the extraction of metal ions into various ILs indicate that these solvents frequently provide extraction efficiencies far greater than those obtained with conventional solvents, other work suggests that they suffer from various drawbacks that could limit their utility as extraction solvents. In this chapter, we examine the viability of ionic liquids as the basis for extraction systems for the separation of actinides and fission products from acidic media and consider approaches by which their limitations may be overcome.

Stepinski, D. C.; Young, B. A.; Jensen, M. P.; Rickert, P. G.; Dzielawa, J. A.; Dilger, A. A.; Rausch, D. J.; Dietz, M. L.; Chemistry

2006-01-01

219

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

220

Evaluation of six decontamination processes on actinide and fission product contamination  

SciTech Connect

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

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

1995-12-31

221

High-Current Superconducting Cyclotron for Accelerator-Driven Subcritical Fission and for Medical Isotope Production  

NASA Astrophysics Data System (ADS)

A 50 MeV, 5mA proton cyclotron is being developed as the injector for a high-current driver for an accelerator-driven subcritical fission power system (ADSMS), and also for production of isotopes for medical physics. Two innovations have made it possible to design a cyclotron capable of >5 mA beam current: strong-focusing of the bunches by quadrupole focusing channels integrated on the pole faces of the sector magnets, and superconducting rf accelerating cavities to provide sufficient energy gain per turn to cleanly separate the orbits. Simulation results will be presented for the beam dynamics of the intense proton bunches during injection, acceleration, and extraction. Key features for both applications will be discussed.

Badgley, Karie; Assadi, Saeed; McIntyre, Peter; Sattarov, Akhdiyor

2011-10-01

222

IBM-1 description of the fission products $^{108,110,112}$Ru  

E-print Network

IBM-1} calculations for the fission products $^{108,110,112}$Ru have been carried out. The even-even isotopes of Ru can be described as transitional nuclei situated between the U(5) (spherical vibrator) and SO(6) ($\\gamma$-unstable rotor) symmetries of the Interacting Boson Model. At first, a Hamiltonian with only one- and two-body terms has been used. Excitation energies and $B$(E2) ratios of gamma transitions have been calculated. A satisfactory agreement has been obtained, with the exception of the odd-even staggering in the quasi-$\\gamma$ bands of $^{110,112}$Ru. The observed pattern is rather similar to the one for a rigid triaxial rotor. A calculation based on a Hamiltonian with three-body terms was able to remove this discrepancy. The relation between the IBM and the triaxial rotor model was also examined.

Stefanescu, I; Jolie, J; Van Isacker, P; Von Brentano, P; Luo, Y X; Zhu, S J; Rasmussen, J O; Hamilton, J H; Ramayya, A V; Che, X L

2007-01-01

223

IBM-1 description of the fission products 108,110,112Ru  

NASA Astrophysics Data System (ADS)

IBM-1 calculations for the fission products 108,110,112Ru have been carried out. The even-even isotopes of Ru can be described as transitional nuclei situated between the U(5) (spherical vibrator) and SO(6) ( ?-unstable rotor) symmetries of the interacting Boson Model. At first, a Hamiltonian with only one- and two-body terms has been used. Excitation energies and B(E2) ratios of gamma transitions have been calculated. A satisfactory agreement has been obtained, with the exception of the odd-even staggering in the quasi- ? bands of 110,112Ru. The observed pattern is rather similar to the one for a rigid triaxial rotor. A calculation based on a Hamiltonian with three-body terms was able to remove this discrepancy. The relation between the IBM and the triaxial rotor model was also examined.

Stefanescu, I.; Gelberg, A.; Jolie, J.; Van Isacker, P.; von Brentano, P.; Luo, Y. X.; Zhu, S. J.; Rasmussen, J. O.; Hamilton, J. H.; Ramayya, A. V.; Che, X. L.

2007-06-01

224

IBM-1 description of the fission products $^{108,110,112}$Ru  

E-print Network

IBM-1} calculations for the fission products $^{108,110,112}$Ru have been carried out. The even-even isotopes of Ru can be described as transitional nuclei situated between the U(5) (spherical vibrator) and SO(6) ($\\gamma$-unstable rotor) symmetries of the Interacting Boson Model. At first, a Hamiltonian with only one- and two-body terms has been used. Excitation energies and $B$(E2) ratios of gamma transitions have been calculated. A satisfactory agreement has been obtained, with the exception of the odd-even staggering in the quasi-$\\gamma$ bands of $^{110,112}$Ru. The observed pattern is rather similar to the one for a rigid triaxial rotor. A calculation based on a Hamiltonian with three-body terms was able to remove this discrepancy. The relation between the IBM and the triaxial rotor model was also examined.

I. Stefanescu; A. Gelberg; J. Jolie; P. Van Isacker; P. Von Brentano; Y. X. Luo; S. J. Zhu; J. O. Rasmussen; J. H. Hamilton; A. V. Ramayya; X. L. Che

2007-06-12

225

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

NASA Astrophysics Data System (ADS)

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

Zhang, Jinsuo

2014-04-01

226

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

227

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

NASA Astrophysics Data System (ADS)

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

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

2005-03-01

228

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

NASA Astrophysics Data System (ADS)

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

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

2007-08-01

229

Nuclear Fission  

Microsoft Academic Search

The potential role of nuclear fission to meet increased future energy demand while reducing greenhouse gas emissions and controlling nuclear proliferation is assessed. The World Energy Council projection for an environmentally driven future is used, which projects deployment of nearly 3 TW(e) of nuclear generation by 2100, with concurrent reduction of global CO2 emissions to one-third of present levels. We

ERICH SCHNEIDER; WILLIAM C. SAILOR

2006-01-01

230

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

231

Experimental determination of the leakage of solid fission products at a temperature simulating an accident at a nuclear power plant  

Microsoft Academic Search

A great deal of attention is now being devoted to the analysis of possible accidents during which overheating of fuel, resulting in the global destruction of the active zone, melting of the active zone, and leakage of fission products into the coolant and beyond the first loop and containment of the nuclear power plant, can occur when cooling is disrupted.

I. V. Zakrzhevskaya; G. V. Momot; A. V. Statkov; A. A. Khrulev; V. P. Shmelev

1993-01-01

232

Experimental determination of the leakage of solid fission products at a temperature simulating an accident at a nuclear power plant  

Microsoft Academic Search

A great deal of attention is now being devoted to the analysis of possible accidents during which overheating of fuel, resulting in the global destruction of the active zone, melting of the active zone, and leakage of fission products into the coolant and beyond the first loop and containment of the nuclear power plant, can occur when cooling is disrupted.

I. V. Zakrzhevskaya; G. V. Momot; A. V. Statkov; A. A. Khrulev; V. P. Shmelev

1994-01-01

233

ION EXCHANGE IN THE ATOMIC ENERGY INDUSTRY WITH PARTICULAR REFERENCE TO ACTINIDE AND FISSION PRODUCT SEPARATION - A REVIEW  

Microsoft Academic Search

This paper reviews some of the uses of ion exchange processes by the nuclear industry for the period April 1978 to April 1983. The topics dealt with are: thorium, protactinium, uranium, neptunium, plutonium, americium, cesium and actinide-lanthanide separations; the higher actinides - Cm, Bk, Cf, Es and Fm; fission products; ion exchange in the geological disposal of radioactive waste. Consideration

I. L. Jenkins

1984-01-01

234

Optimization and evaluation of mixed-bed chemisorbents for extracting fission and activation products from marine and fresh waters  

Microsoft Academic Search

Chemically selective chemisorbents are needed to monitor natural and engineered waters for anthropogenic releases of stable and radioactive contaminants. Here, a number of individual and mixtures of chemisorbents were investigated for their ability to extract select fission and activation product elements from marine and coastal waters, including Co, Zr, Ru, Ag, Te, Sb, Ba, Cs, Ce, Eu, Pa, Np, and

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

2011-01-01

235

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

236

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions  

Microsoft Academic Search

For the assessment of ? -ray doses from short-lived fission products (FPs) under criticality accident conditions, ? -ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of ? -ray from

Hiroshi YANAGISAWA; Akio OHNO; Eijyu AIZAWA

2002-01-01

237

Measurement of gas production of microorganisms  

Microsoft Academic Search

A simple apparatus and method is disclosed for measuring gas production by microorganisms using a pressure transducer to sense pressure buildup by members of the Enterobacteriaceae group of bacteria. The test system consists of a 5.0 psid pressure transducer and a pressure equalizer valve attached to the metal cap of a 20 x 150 mm test tube. Gas pressure is

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

1975-01-01

238

Monthly Natural Gas Gross Production Report  

EIA Publications

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

2014-01-01

239

Lake Erie gas production posed unique problems  

SciTech Connect

Thick, soft bottom sediments plus the presence of H/sub 2/S, CO/sub 2/ and hydrates required application of special gas production techniques. Wellheads needed modification, flexible flowline connections were used and potential dangers from sour gas were handled with inhibitors and a safety shut-down system.

Sangster, R.B.

1981-09-01

240

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

241

Bio-Gas Production from Alligator Weeds.  

National Technical Information Service (NTIS)

Laboratory experiments were conducted to study the effect of temperature, sample preparation, reducing agents, light intensity and pH of the media, on bio-gas and methane production from the microbial anaerobic decomposition of alligator weeds (Alternanth...

A. Latif

1976-01-01

242

Production Trends of Shale Gas Wells  

E-print Network

PRODUCTION TRENDS OF SHALE GAS WELLS A Thesis by WAQAR ALI KHAN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE December... 2008 Major Subject: Petroleum Engineering PRODUCTION TRENDS OF SHALE GAS WELLS A Thesis by WAQAR ALI KHAN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements...

Khan, Waqar A.

2010-01-14

243

A ROOT-based analysis tool for measurements of neutron-induced fission products at the IGISOL facility  

NASA Astrophysics Data System (ADS)

For the sustainable development of nuclear energy, the handling of used nuclear fuel is a key issue. Innovative fuel cycles are being developed for the transmutation of minor actinides and long-lived fission products. In view of these developments, accurate knowledge of the fuel inventory is necessary. The IGISOL facility with JYFLTRAP, at the accelerator laboratory of the University of Jyvskyl, will be used to measure independent fission yield distributions from neutron-induced fission on different actinides. In this paper, an analysis tool is developed, using the CERN-based ROOT Data Analysis Framework, with the objective of performing full data analysis within the same code. The analysis tool is currently being tested on the data from measurements with 25 MeV protons on a 232Th target, and some preliminary results are presented.

Mattera, A.; Gorelov, D.; Lantz, M.; Lourdel, B.; Penttil, H.; Pomp, S.; Ryzhov, I.

2012-10-01

244

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

245

Partitioning of fission products from irradiated nitride fuel using inductive vaporization  

SciTech Connect

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

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

2013-07-01

246

Radioactive Ion Beam Production from the Fission of Thorium Oxide Targets  

NASA Astrophysics Data System (ADS)

Hollifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory is one of the few facilities in the world that provides radioactive ion beams (RIB), crucial for nuclear astrophysics, nuclear structure, and stewardship science. Neutron-rich beams are produced by proton-induced nuclear fission of actinide compounds such as uranium carbide or thorium oxide. The goal of this project has two folds. First, compare the beam yield produced from both a low density and a high-density ThO2 target. Second, find the relation the 40 MeV proton beam that drives the RIB production is fully stopped in the high density, 8 g/cm^3 ThO2, but not in the low-density 0.8 g/cm^3 ThO^2. The low-density target does not use all of the beam intensity. In this particular experiment, the production yields from 40MeV and 30MeV protons have been measured on the low-density target. The comparison of the calculated production yields of 40 MeV and 30 MeV protons shows a factor of two between these different energies. The experiment was conducted using an on-line mass separator, and specific masses of the RIB were collected onto a tape. This allows a direct comparison of the low and high density ThO2 target. Release data from the high and low-density targets will be shown and discussed.

Armagan, Hakan; Carter, H. K.; Stracener, D. W.; Spejewski, E. H.; Kronenberg, A.

2007-04-01

247

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

248

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

Microsoft Academic Search

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

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

1980-01-01

249

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

SciTech Connect

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

Gauntt, Randall O.

2010-04-01

250

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

NASA Astrophysics Data System (ADS)

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 U-Mo 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 capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

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

2009-04-01

251

Low enriched uranium foil plate target for the production of fission Molybdenum99 in Pakistan Research Reactor1  

Microsoft Academic Search

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

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

2009-01-01

252

MASS DISTRIBUTION OF FISSION PRODUCTS PRODUCED BY IRRADIATION OF GOLD AND URANIUM BY NITROGEN IONS  

Microsoft Academic Search

The mass spectrum of fission fragments produced by irradiating thick ; targets of gold and uranium with 115 Mev nitrogen ions was investigated. ; Fourteen different elements were separated from the irradiated targets. The mass ; distribution curve for fission fragments produced by irradiating gold has the ; form of a narrow peak with a haif width of about 20

N. I. Tarantin; Yu. B. Gerlit; L. I. Guseva; B. F. Myasoedov; Fillippova; G. N. K. V. f Flerov

1958-01-01

253

Greenhouse gas budgets of crop production current  

E-print Network

land use change 16 2.7 Indirect emissions from crop production 16 2.7.1 Emissions from fertilizer and increased cover 42 4.8.2 Crop selection and rotation 42 4.9 Mitigation potential in rice production 43 4 negotiations on climate change. It provides an up-to-date state of the scientific knowledge on greenhouse gas

Levi, Ran

254

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

255

Evolution of gas saturation and relative permeability during gas production from hydrate-bearing sediments: Gas invasion vs. gas nucleation  

NASA Astrophysics Data System (ADS)

and both gas and water permeabilities change as a function of gas saturation. Typical trends established in the discipline of unsaturated soil behavior are used when simulating gas production from hydrate-bearing sediments. However, the evolution of gas saturation and water drainage in gas invasion (i.e., classical soil behavior) and gas nucleation (i.e., gas production) is inherently different: micromodel experimental results show that gas invasion forms a continuous flow path while gas nucleation forms isolated gas clusters. Complementary simulations conducted using tube networks explore the implications of the two different desaturation processes. In spite of their distinct morphological differences in fluid displacement, numerical results show that the computed capillarity-saturation curves are very similar in gas invasion and nucleation (the gas-water interface confronts similar pore throat size distribution in both cases); the relative water permeability trends are similar (the mean free path for water flow is not affected by the topology of the gas phase); and the relative gas permeability is slightly lower in nucleation (delayed percolation of initially isolated gas-filled pores that do not contribute to gas conductivity). Models developed for unsaturated sediments can be used for reservoir simulation in the context of gas production from hydrate-bearing sediments, with minor adjustments to accommodate a lower gas invasion pressure Po and a higher gas percolation threshold.

Jang, Jaewon; Santamarina, J. Carlos

2014-01-01

256

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

NASA Astrophysics Data System (ADS)

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

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

2014-10-01

257

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

258

State taxation of oil and gas production  

SciTech Connect

Detailed information on state oil and gas production taxes reports the level of taxes paid in recent years and demonstrates the effect of prices, tax rates, and levels of production on these amounts. The report also describes the implementation of the production taxes, including details of who pays them, at what rate, how tax revenues are distributed, and other features. The states use these taxes as a source of general funds, and allocate some fraction to local governments if the taxes preempt local property taxes. Historically, movements in oil and gas prices had the most impact on tax collections, but economists note that the five new production taxes and 10 increases in tax rates since 1979 reduce the incentives for investment, which leads to decreased future production. 1 figure, 3 tables.

Smith, F.H.

1985-01-01

259

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

260

Methane hydrate gas production by thermal stimulation  

SciTech Connect

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 were made to determine the importance of a number of variables, including porosity, bed thickness, injection temperature, and fracture length. These studies indicate that the hydrate-filled porosity should be at least 15%, reservoir thickness should be about 25 ft or more, and well spacing should be fairly large (maybe 40 acres/well), if possible. Injection temperatures should probably be between 150 and 250/sup 0/F to achieve an acceptable balance between high heat losses and unrealistically high injection rates. Numerous important questions about hydrate gas production remain unanswered.

McGuire, P.L.

1981-01-01

261

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

262

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

SciTech Connect

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

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

2014-01-01

263

Electrochemical separation of actinides and fission products in molten salt electrolyte  

SciTech Connect

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

Gay, R. L.; Grantham, L. F.; Fusselman, S. P.; Grimmett, D. L.; Roy, J. J. [Rockwell International/Rocketdyne Division Canoga Park, California 91309-7922 (United States)

1995-09-15

264

Electrochemical separation of actinides and fission products in molten salt electrolyte  

NASA Astrophysics Data System (ADS)

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

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

1995-09-01

265

Studies on the recovery of /sup 99/Mo from uranium fission products  

SciTech Connect

A study for the recovery of carrier-free and high-specific-activity /sup 99/Mo (parent nuclide of radiopharmaceutical /sup 99m/Tc) from uranium fission products has been undertaken. A liquid-liquid extraction system consisting of a new, stable, and almost odor-free high-molecular-weight analogue of the pyridine series, i.e., 4-(5-nonyl)pyridine, whose analytical potentials have already been established, has been used as a solvent extraction reagent. Molybdenum (VI) is extracted with extremely high distribution coefficients (> 10/sup 3/), with only 0.1-M solution of the solvent in a carrier diluent, benzene, from 0.1-M aqueous sulfuric acid in 0.1- to 1.0-M potassium thiocyanate. The system requires a very small concentration of the complexing agent and the supporting acid and does not require the use of salting-out agents, which generally add to the radioactive waste disposal problem. Optimization of the parameters has been accomplished by extensive investigations of different parameters affecting the distribution equilibria of the metal. The extraction mechanism has been investigated through slope analysis and loading ratio data. It has been found that two molecules of the solvent are involved per complex of the extracted metal, which is most probably MoO/sub 2/ (SCN)/sub 2/ and is independent of the nature of the anions of the supporting acids. Back extraction of the metal can easily be accomplished with 1.0-M nitric acid.

Ejaz, M.; Mammoon, A.M.

1987-01-01

266

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

E-print Network

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

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

2004-01-01

267

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

E-print Network

We present measurements of airborne fission products in Chapel Hill, NC, USA, from 62 days 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 I-131 and Cs-137 were measured with maximum activities of 4.2 +/- 0.6 mBq/m^3 and 0.42 +/- 0.07 mBq/m^3 respectively. Additional activity from I-131, I-132, Cs-134, Cs-136, Cs-137 and Te-132 were measured in the same air filters using a low-background HPGe detector at the Kimballton Underground Research Facility (KURF).

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

2011-11-17

268

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

269

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

SciTech Connect

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

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

2007-11-15

270

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

Microsoft Academic Search

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

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

2011-01-01

271

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

Microsoft Academic Search

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

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

2011-01-01

272

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

Microsoft Academic Search

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

A. Gruel; P. Leconte; D. Bernard

2010-01-01

273

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

274

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

275

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

276

Bio-gas production from alligator weeds  

NASA Technical Reports Server (NTRS)

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

Latif, A.

1976-01-01

277

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

SciTech Connect

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

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

1980-04-01

278

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

279

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

280

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

SciTech Connect

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

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

2010-11-01

281

Shale Gas Production: Potential versus Actual GHG Emissions  

E-print Network

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

282

Conceptual Analysis of the Power Production of Fission Electric Cell Reactors  

SciTech Connect

The United States Department of Energy, Nuclear Energy Research Initiative (NERI) Direct Energy Conversion project has as its goal the development of direct energy conversion (DEC) processes suitable for commercial development. DEC is defined as any fission process that returns usable energy with no intermediate thermal process. This project includes the study of the fission electric cell (FEC). In the FEC, fission fragments exit the fuel element cathode and are collected by the cell anode. Previous work [1] has shown the potential of FECs, with theoretical efficiencies up to 60%. Inspection of this work indicates the need for additional system modeling prior to any conclusions regarding the final FEC reactor configuration. This paper builds on the previous work and outlines the development of models to facilitate design decisions. The models address criticality, design life, reactor configuration, and current-voltage characteristics. In addition, this paper proposes future work to complete the design model. (authors)

King, Donald; Rochau, Gary; Morrow, Charles; Cash, Jamie; Seidel, David; Slutz, Stephen [Sandia National Laboratories (United States)

2002-07-01

283

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

SciTech Connect

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

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

2009-02-01

284

Production of noble gas isotopes by proton-induced reactions on lead  

NASA Astrophysics Data System (ADS)

We measured integral thin target cross sections for the proton-induced production of He-, Ne-, Ar-, Kr- and Xe-isotopes from lead from the respective reaction threshold up to 2.6 GeV. The production of noble gas isotopes from lead is of special importance for design studies of accelerator driven nuclear reactors and/or energy amplifiers. For all experiments with proton energies above 200 MeV a new mini-stack approach was used instead of the stacked-foil technique in order to minimise the influences of secondary particles on the residual nuclide production. About 420 cross sections for 23 nuclear reactions were determined. The phenomenology of the determined excitation functions enables us to distinguish between the different reaction modes fragmentation, hot and cold symmetric fission, asymmetric fission and deep spallation. Cross sections for the production of 21Ne and 38Ar measured below 100 MeV and 200 MeV, respectively, enable us to study nuclide production below the nominal Coulomb-barrier. The experimental data are compared to results from the theoretical nuclear model code INCL4/ABLA. While the model describes the production of 4He reasonably well, it underestimates the cross sections for Ne- and Ar-isotopes produced via deep spallation and/or multifragmentation by up to two orders of magnitude. For the Kr- and Xe-isotopes the agreement between modelled and measured data strongly depends on the reaction mechanisms. While INCL4/ABLA describes the production of n-poor Kr-isotopes via hot-symmetric fission and the production of Xe-isotopes via asymmetric fission reasonably well, i.e. within a factor of 2, the discrepancies between modelled and measured cross sections for the n-rich Kr-isotopes produced via cold symmetric fission are significantly larger. For the Xe-isotopes produced via spallation, i.e. at energies higher than about 600 MeV, the model completely fails to describe the experimental data. Therefore, the comparison of measured and modelled thin target cross sections clearly indicates that experimental data are still needed because the predictive power of nuclear model codes, though permanently improving, does still not allow to reliably predict the cross sections needed for most applications and irradiation experiments remain indispensable.

Leya, I.; Wieler, R.; David, J.-C.; Leray, S.; Donadille, L.; Cugnon, J.; Michel, R.

2005-02-01

285

Gas production in the MEGAPIE spallation target  

SciTech Connect

The Megawatt Pilot Experiment (MEGAPIE) project was started in 2000 to design, build and operate a liquid Lead-Bismuth Eutectic (LBE) spallation neutron target at the power level of 1 MW. The target was irradiated for four months in 2006 at the Paul Scherrer Inst. in Switzerland. Gas samples were extracted in various phases of operation and analyzed by {gamma} spectroscopy leading to the determination of the main radioactive isotopes released from the LBE. Comparison with calculations performed using several validated codes (MCNPX2.5.0/CINDER'90, FLUKA/ORIHET and SNT) yields the ratio between simulated in-target isotope production rates and experimental amount released at any given time. This work underlines the weak points of spallation models for some released isotopes. Also, results provide relevant information for safety and radioprotection in an Accelerator Driven System (ADS) and more particularly for the gas management in a spallation target dedicated to neutron production facilities. (authors)

Thiolliere, N. [SUBATECH, EMN-IN2P3/CNRS-Universite, Nantes, F-44307 (France); Zanini, L. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); David, J. C. [CEA Saclay, Irfu/SPhN, 91191 Gif Sur Yvette (France); Eikenberg, J. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Guertin, A. [SUBATECH, EMN-IN2P3/CNRS-Universite, Nantes, F-44307 (France); Konobeyev, A. Y. [Institut fuer Reaktorsicherheit, FZK GmbH, 76021 Karlsruhe (Germany); Lemaire, S. [CEA Bruyeres-le-Chatel, DAM Ile de France, 91297 Arpajon Cedex (France); Panebianco, S. [CEA Saclay, Irfu/SPhN, 91191 Gif Sur Yvette (France)

2011-07-01

286

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

NASA Astrophysics Data System (ADS)

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

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

287

Modification of a novel macroporous silica-based crown ether impregnated polymeric composite with 1-dodecanol and its adsorption for some fission and non-fission products contained in high level liquid waste  

Microsoft Academic Search

A novel macroporous silica-based chelating polymeric composite, DtDo\\/SiO2P, was synthesized by molecular modification of 4,4?,(5?)-di-(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) with a long carbon chain organic compound 1-dodecanol. It was performed through impregnation and immobilization of DtBuCH18C6 and 1-dodecanol molecules into the pores of the SiO2P particles. The adsorption of a few fission and non-fission product elements Sr(II), Ba(II), Cs(I), Ru(III), Mo(VI), Na(I), K(I),

Anyun Zhang; Weihong Wang; Zhifang Chai; Etsushu Kuraoka

2008-01-01

288

Shale Gas Production: Potential versus Actual GHG Emissions  

E-print Network

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

O'Sullivan, Francis

289

Fission fusion hybrids- recent progress  

NASA Astrophysics Data System (ADS)

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

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

2012-03-01

290

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

E-print Network

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

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

2014-05-14

291

$gamma$-RAYS FROM SHORT-LIVED FISSION PRODUCTS OF ²³⁵U AND ²³⁹Pu  

Microsoft Academic Search

Results are given of some measurements of the spectral composition of ; the gamma rays emitted by U²³⁵ fission products in the time interval of ; 1.5 to 5 sec after fission. The absorption coefficient, using Cu absorbers, was ; 0.048 plus or minus 0.005 cm²\\/g amd was 0.046 plus or minus 0.009 cm\\/sup ; 2\\/\\/g for Al absorbers. In

O. I. Leipunsky; V. N. Saharov; V. I. Tereschenko

1957-01-01

292

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

SciTech Connect

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

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

2010-09-23

293

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

E-print Network

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

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

2011-01-01

294

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

SciTech Connect

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

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

2006-07-01

295

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

NASA Astrophysics Data System (ADS)

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

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

2014-04-01

296

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

PubMed

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

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

2014-01-01

297

Production of noble gas isotopes by proton-induced reactions on lead and bismuth  

NASA Astrophysics Data System (ADS)

We measured integral thin target cross-sections for the proton-induced production of He-, Ne-, Ar-, Kr-, and Xe-isotopes from lead and bismuth from the respective reaction threshold up to 2.6 GeV. The production of noble gas isotopes from lead and bismuth is of special importance for design studies of accelerator driven nuclear reactors and/or energy amplifiers. For all experiments with proton energies above 200 MeV a new mini-stack approach was used instead of the stacked-foil technique in order to minimise influences of secondary particles. The phenomenology of the determined excitation functions enables us to distinguish between the different reaction modes fragmentation, hot and cold symmetric fission, asymmetric fission, and deep spallation. For lead more than 420 cross-sections for 23 nuclear reactions have been measured. While the lead data have already been published, here we present first results for the production of noble gas isotopes from bismuth. The experimental data are compared to results from the theoretical nuclear model code INCL4/ABLA. This comparison clearly indicates that experimental data are still needed because the predictive power of nuclear model codes, though permanently improving, does still not allow to reliably predict the cross-sections needed for most applications and irradiation experiments remain indispensable.

Leya, I.; Wieler, R.; David, J.-C.; Leray, S.; Donadille, L.; Cugnon, J.; Michel, R.

2006-06-01

298

Landfill course: managing gas and leachate production on landfills  

Microsoft Academic Search

Controlling gas and leachate production is a primary objective of sound landfill management. Gases, composed mainly of carbon dioxide and methane, are formed during the decomposition of solid wastes. Leachate forms as water passes through the refuse, dissolving out chemicals. Three basic methods for controlling gas and leachate production are presented: managing production; directing gas or leachate movement; and treating

Reinfl

1977-01-01

299

Lifetime measurements of excited levels in prompt fission products of 252Cf  

Microsoft Academic Search

Lifetimes in the range 10-11 to 10-9 s of prompt gamma rays emitted from the fission fragments of 252Cf were measured using a recoil distance method. A 252Cf source was deposited on a stretched Ni foil and placed in a plunger device, the recoil direction of the studied fragments being determined by the detection of the complementary fragment. The lifetime

G. Mamane; E. Cheifetz; E. Dafni; A. Zemel; J. B. Wilhelmy

1986-01-01

300

78 FR 59650 - Subzone 9F, Authorization of Production Activity, The Gas Company, LLC dba Hawai'i Gas...  

Federal Register 2010, 2011, 2012, 2013

...Subzone 9F, Authorization of Production Activity, The Gas Company, LLC dba Hawai'i Gas, (Synthetic Natural Gas), Kapolei, Hawaii On May 22, 2013, The Gas Company, LLC dba Hawai'i Gas submitted a notification of proposed production...

2013-09-27

301

Bio Gas Oil Production from Waste Lard  

PubMed Central

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

Hancsok, Jeno; Baladincz, Peter; Kasza, Tamas; Kovacs, Sandor; Toth, Csaba; Varga, Zoltan

2011-01-01

302

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

303

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

SciTech Connect

Fission-product decay energy-release rates were measured for thermal-neutron fission of /sup 239/Pu. Samples of mass 1 and 5 ..mu..g were irradiated for 1 to 100 s using the fast pneumatic-tube facility at the Oak Ridge Research Reactor. The resulting beta- and gamma-ray emissions were separately counted for times-after-fission between 2 and 14,000 s to yield spectral distributions N(E/sub ..gamma../) vs E/sub ..gamma../ and N(E/sub ..beta../) vs E/sub ..beta../. The gamma-ray spectra were obtained by use of a NaI detector, and the beta-ray spectra were obtained by use of an NE-110 detector with an anticoincidence mantle. The raw data were unfolded to provide spectral distributions of moderate resolution. These distributions are given in graphical and tabular form as differential spectral intensity I(E) (MeV/sup -1/ fission/sup -1/) averaged over gamma-ray energy intervals ranging from 10 keV for E/sub ..gamma../ < 0.18 MeV to 100 keV for E/sub ..gamma../ > 6.8 MeV, and beta-ray energy intervals ranging from 20 keV for E/sub ..beta../ < 0.25 MeV to 160 keV for E/sub ..beta../ > 6.4 MeV. Counting-time intervals ranged from 1 s for times-after-fission (t/sub w/) < 6 s to 4000 s for t/sub w/ approx. 10/sup 4/ s. For comparisons the graphical representations show calculated spectra obtained by use of the CINDER-10 summation code and the ENDF/B-IV fission yield and decay scheme data base. 90 figures, 86 tables.

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

1980-01-01

304

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

305

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

NASA Astrophysics Data System (ADS)

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

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

2012-11-01

306

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

SciTech Connect

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides {sup 235,238}U and {sup 239}Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on {sup 239}Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication 'ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,' Nuclear Data Sheets 107, 2931 (2006).

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

2011-12-01

307

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

308

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

309

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

SciTech Connect

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

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

2007-12-01

310

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.

311

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

312

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

NASA Astrophysics Data System (ADS)

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

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

313

A hybrid economicengineering model for natural gas production  

Microsoft Academic Search

An optimal control model which generalizes the traditional economic theory of exhaustible resource production is developed and applied to natural gas wells. These generalizations, which are empirically relevant for the natural gas resources we analyze, allow (1) decreasing marginal production costs, (2) physical bounds on periodic production and (3) interdependencies between the stock of the resource, the periodic production bounds,

Janie M. Chermak; James Crafton; Suzanne M. Norquist; Robert H. Patrick

1999-01-01

314

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

315

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

NASA Astrophysics Data System (ADS)

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 been introduced simulates anisotropic deformation of the uranium-zirconium alloy fuel. The models included in the ALFUS are thought reasonable and consistent with knowledge obtained from the irradiation test results. When the fuel slug swells out and comes into contact with the cladding, compressive stress is produced in the slug and decreases volume of the open pore if it has been formed. This is an essential process to keep fuel-cladding mechanical interaction (FCMI) small in case of lower smear density fuel. Analyses with the ALFUS indicate a significant level of FCMI in case of higher than 80% smear density fuel.

Ogata, Takanari; Kinoshita, Motoyasu; Saito, Hiroaki; Yokoo, Takeshi

1996-06-01

316

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

PubMed Central

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

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

2014-01-01

317

Spontaneous and induced emission of XeCl* excimer molecules under pumping of Xe - CCl4 and Ar - Xe - CCl4 gas mixtures with a low CCl4 content by fast electrons and uranium fission fragments  

NASA Astrophysics Data System (ADS)

The spontaneous and induced emission of XeCl* excimer molecules upon excitation of Xe - CCl4 and Ar - Xe - CCl4 gas mixtures with a low CCl4 content by high-energy charged particles [a pulsed high-energy electron beam and products of neutron nuclear reaction 235U(n, f)] has been experimentally studied. The electron energy was 150 keV, and the pump current pulse duration and amplitude were 5 ns and 5 A, respectively. The energy of fission fragments did not exceed 100 MeV, the duration of the neutron pump pulse was 200 ?s, and the specific power contribution to the gas was about 300 W cm-3. Electron beam pumping in a cell 4 cm long with a cavity having an output mirror transmittance of 2.7% gives rise to lasing on the B ? X transition in the XeCl* molecule (? = 308 nm) with a gain ? = 0.0085 cm-1 and fluorescence efficiency ? ? 10%. Pumping by fission fragments in a 250-cm-long cell with a cavity formed by a highly reflecting mirror and a quartz window implements amplified spontaneous emission (ASE) with an output power of 40 - 50 kW sr-1 and a base ASE pulse duration of ~200 ms.

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

2013-11-01

318

Measurements of Methane Emissions at Natural Gas Production Sites  

E-print Network

· Sponsors were an environmental group and nine natural gas producers ­ Environmental Defense Fund (EDFMeasurements of Methane Emissions at Natural Gas Production Sites in the United States #12;Why = 21 #12;Need for Study · Estimates of methane emissions from natural gas production , from academic

Lightsey, Glenn

319

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

E-print Network

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

American Society for Testing and Materials. Philadelphia

2005-01-01

320

Ground movements associated with gas hydrate production  

SciTech Connect

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

Siriwardane, H.J.

1992-10-01

321

Corrosion inhibition in oil and gas production  

SciTech Connect

This paper discusses practical aspects of the design and implementation of a corrosion inhibition program for oil and gas production, including choice of system, inhibitor testing and selection, performance monitoring, and inhibition problems. Corrosion inhibition is used to ensure safe operations and to improve profits. This paper focuses on the economic aspects of inhibition. The type of failures discussed here involve loss of production or assets, but not catastrophic failures (eg, well blow-out) where safety or environmental factors predominate. Corrosion failures must be economically significant to justify the implementation of a downhole corrosion inhibition program. Inhibition is one of several alternative methods for controlling corrosion. The choice of method is driven by both economical and technical considerations. The criteria include: technical competence, feasibility of implementation, compatibility with the rest of the production system, and initial and maintenance costs. The design of a corrosion inhibition program involves several steps from inhibitor selection to performance monitoring. Each step of the process must be carefully addressed to ensure the success of the program.

Kapusta, S.D. [Shell Western E and P Inc., Houston, TX (United States); Place, M.C. [Project Associates, Metairie, LA (United States)

1994-12-31

322

Spin Assignments, Mixing Ratios, and g-FACTORS in Neutron Rich 252Cf Fission Products  

NASA Astrophysics Data System (ADS)

We present a new technique for measuring angular correlations between ?-rays emitted by the fragments from the spontaneous fission of 252Cf and measured with Gammasphere. For states with short lifetimes (?10ps), these correlations can be used to determine the spin and parity of unknown levels. For states with long lifetimes, the technique can be used to determine the g-factor of the level in question by measuring the attenuation of the correlation caused by rotation of the nucleus about the randomly oriented domains in an un-magnetized iron foil. Applying our new method to our set of triple coincidence data, we have been able to assign spins to new levels in 108,110,112Ru. Mixing ratio and g-factor measurements are also discussed.

Goodin, C.; Daniel, A. V.; Li, K.; Ramayya, A. V.; Hwang, J. K.; Hamilton, J. H.; Stone, N. J.; Stone, J. R.; Rasmussen, J. O.

2008-08-01

323

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

SciTech Connect

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

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

2006-07-01

324

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

SciTech Connect

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

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

2010-04-01

325

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

E-print Network

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

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

2011-03-24

326

Numerical solution for the nonlinear diffusion equation describing the transport of volatile fission products in nuclear fuels  

SciTech Connect

This paper reports that to describe the transport of volatile fission products along grain boundaries in nuclear fuels, a nonlinear diffusion equation must be used. Analytic solutions exist for the steady-state case, but the equation seems to be intractable when time dependence is included. A simple implicit numerical method has been developed that can guarantee a convergent stable solution when there is a central void. If there is no void, the method always yields a solution. There is perfect agreement between the analytic and numerical solutions for the steady state, and the method developed here offers significant advantages over other methods of solution. This basic model can be used in nuclear fuel performance studies.

O'Carroll, C.; Lassmann, K. (Commission of the European Communities, Joint Research Center, Inst. for Transuranium Elements, Postfach 2340, D-7500 Karlsruhe (DE))

1992-08-01

327

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

PubMed Central

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

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

2014-01-01

328

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

Microsoft Academic Search

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

S. Frank

2009-01-01

329

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

NASA Astrophysics Data System (ADS)

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

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

2014-11-01

330

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

331

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

E-print Network

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

Mengal, Salman Akram

2010-10-12

332

Electromagnetic-induced fission of 238U projectile fragments, a test case for the production of spherical super-heavy nuclei  

NASA Astrophysics Data System (ADS)

Isotopic series of 58 neutron-deficient secondary projectiles ( 205,206At, 205-209Rn, 208-212,217,218Fr, 211-223Ra, 215-226Ac, 221-229Th, 226-231Pa, 231-234U) were produced by projectile fragmentation using a 1 A GeV 238U beam. Cross sections of fission induced by nuclear and electromagnetic interactions in a secondary lead target were measured. They were found to vary smoothly as a function of proton and neutron number of the fissioning system, also for nuclei with large ground-state shell effects near the 126-neutron shell. No stabilization against fission was observed for these nuclei at low excitation energies. Consequences for the expectations on the production cross sections of super-heavy nuclei are discussed.

Heinz, A.; Schmidt, K.-H.; Junghans, A. R.; Armbruster, P.; Benlliure, J.; Bckstiegel, C.; Clerc, H.-G.; Grewe, A.; de Jong, M.; Mller, J.; Pftzner, M.; Steinhuser, S.; Voss, B.

2003-01-01

333

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

334

Production of biodiesel using expanded gas solvents  

SciTech Connect

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

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

2009-04-07

335

Windowless gas targets for neutron production  

NASA Astrophysics Data System (ADS)

A windowless deuterium gas target has been constructed for high yield production of either monoenergetic or white fast neutrons. The operation of this target has been demonstrated on a 900 keV deuteron accelerator. The target is capable of operation at 100 mbar target pressure, and can admit a low duty factor beam of 5 mm transverse extent. The target employs an intermittent valve arrangement to reduce the flow rates in the higher pressure stages of a differentially pumped vacuum system. This valve allows operation at much greater target pressures for low duty factor beams than would otherwise be the case. Neutron yield measurements validated the functionality of the target system. This target will make possible considerable advances in methods of non-destructive testing and evaluation which employ fast neutrons, whether mono-energetic or otherwise. It is further suited to use as a thermal neutron source, with the addition of an appropriate moderator. The development of this target system has not only provided a functioning and valuable piece of equipment for use in further research, but has also investigated the technological limitations and functional requirements of implementing such a system in a practical setting. (Copies available exclusively from MIT Libraries, Rm. 14- 0551, Cambridge, MA 2139-4307. Ph. 617-253-5668; Fax 617- 253-1690.)

Iverson, Erik B.

336

A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE  

SciTech Connect

A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

Lewis, B.J.; Husain, A.

2003-02-27

337

30 CFR 202.550 - How do I determine the royalty due on gas production?  

Code of Federal Regulations, 2010 CFR

...determine the royalty due on gas production? 202.550 Section 202...REVENUE MANAGEMENT ROYALTIES Gas Production From Indian Leases 202.550...determine the royalty due on gas production? If you produce gas...

2010-07-01

338

Gas production and migration in landfills and geological materials.  

PubMed

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. PMID:11695741

Nastev, M; Therrien, R; Lefebvre, R; Glinas, P

2001-11-01

339

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

340

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

341

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

Microsoft Academic Search

The use of a low-enriched uranium (LEU) fuel target was examined for the feasibility of ⁹⁹Mo production in a High-flux Advanced Neutron Application Reactor (HANARO). Uncertainty analysis was done with respect to the ⁹⁹Mo yield ratio, ²³⁹Pu yield ratio, annual production rate, and decontamination requirement. Validity of a coupled code system, MCNP\\/ORIGEN2, was evaluated to estimate reliable isotopic number densities

Dong-Keun Cho; Myung-Hyun Kim

2003-01-01

342

78 FR 59632 - Oil and Gas and Sulphur Operations on the Outer Continental Shelf-Oil and Gas Production Safety...  

Federal Register 2010, 2011, 2012, 2013

...DAQ000 EEEE500000] RIN 1014-AA10 Oil and Gas and Sulphur Operations on the Outer Continental Shelf--Oil and Gas Production Safety Systems AGENCY...20170-4817. Please reference ``Oil and Gas Production Safety Systems,...

2013-09-27

343

Gas well operation with liquid production  

SciTech Connect

Prediction of liquid loading in gas wells is discussed in terms of intersecting tubing or system performance curves with IPR curves and by using a more simplified critical velocity relationship. Different methods of liquid removal are discussed including such methods as intermittent lift, plunger lift, use of foam, gas lift, and rod, jet, and electric submersible pumps. Advantages, disadvantages, and techniques for design and application of the methods of liquid removal are discussed.

Lea, J.F.; Tighe, R.E.

1983-02-01

344

Nuclear Fission  

Microsoft Academic Search

Nuclear fission is a process in which a heavy nucleus splits into two much lighter nuclei. For some very unstable nuclei fission\\u000a can happen spontaneously, but that is a very rare event. Usually, the process is induced by the excitation of the nuclei by\\u000a bombarding them with particles or with gamma rays. Heavy nuclei have a greater neutron\\/proton ratio than

Hanne Andersen

345

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

Code of Federal Regulations, 2012 CFR

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

2012-07-01

346

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

347

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

NASA Astrophysics Data System (ADS)

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

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

2009-09-01

348

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

349

Experimental Measurements of Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air  

SciTech Connect

Experimental measurements of delayed fission-product gamma-ray transmission through low-enriched UO{sub 2} fuel pin lattices in an air medium were conducted at the Rensselaer Polytechnic Institute Reactor Critical Facility (RCF). The RCF core consists of excess Special Power Excursion Reactor Test (SPERT) fuel pins, enriched to 4.81 weight percent {sup 235}U, clad in stainless steel. An experimental apparatus was constructed to hold various arrangements of fuel pin lattices. The arrangements consisted of a single activated source pin taken from the reactor core surrounded by inactive fuel pins in an air medium. A sodium-iodide detector and gamma-ray spectroscopy system was used to generate a pulse-height spectrum of the gamma-ray radiation for detector positions outside the lattice. The change in radiation intensity as the detector is rotated about the vertical axis of the lattice, the ''channeling effect,'' was measured. Experimental measurements of the channeling effect were performed for six arrangements; 3 x 3, 5 x 5, and 7 x 7 lattices, with both the comer position and center position containing the activated pin. The results of the measurements demonstrate that the gamma-ray radiation intensity can vary widely, as a function of angle, relative to the vertical axis of the lattice.

T Trumbull; D Harris

2005-01-26

350

Design, construction, and testing of a 2000°C furnace and fission product collection system  

Microsoft Academic Search

An induction furnace, capable of operation at 2000°C in steam, was developed to conduct product release tests. The test specimen and steam atmosphere are contained in a stabilized ZrO furnace tube, which is heated by a concentric susceptor of either tungsten or graphite. A two-color optical pyrometer and high-temperature thermocouples are used for temperature measurement. The furnace has operated reliably

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

1984-01-01

351

Computer model for production forecasting of oil and gas  

SciTech Connect

In nature, gas generally is found in combination with oil. To make the situation even more complicated, the gas-oil ratio of the crude may vary from well to well, and from time to time within the same well. Without a computer program, the production manager must decide arbitrarily, from past experience and by trial and error, how much crude and gas should be produced by each gas-oil separator (GOSP) to satisfy all demands and constraints. This approach involves numerous staff meetings. The GOSP model described was developed as a means of finding production guidelines that could either maximize or minimize gas production while maintaining a fixed amount of crude output, and vice versa. Each month the guidelines for this model have to be adjusted to meet changing market requirements, surface facility limitations, reservoir constraints, scheduled equipment shutdowns, and other variables.

Hahn, K.W.; Turaiki, S.A.; Al-Mishari, A.S.

1983-07-01

352

Process for production desulfurized of synthesis gas  

DOEpatents

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

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

1993-01-01

353

US/FRG umbrella agreement for cooperation in GCR development. Fuel, fission products, and graphite subprogram. Quarterly status report, January 1, 1983-March 31, 1983  

SciTech Connect

This report describes the status of the cooperative work being performed in the Fuel, Fission Product, and Graphite Subprogram under the HTR-Implementing Agreement of the United States/Federal Republic of Germany Umbrella Agreement for Cooperation in GCR Development. The status is described relative to the commitments in the Subprogram Plan for Fuel, Fission Products, and Graphite, Revision 5, April 1982, and Revision 6, February 1983. The work described was performed during the period January 1 through March 31, 1983 in the HTGR Base Technology Program at Oak Ridge National Laboratory, the HTGR Fuel and Plant Technology Programs at GA Technologies Inc. (GA), and the Project HTR-Brennstoffkreislauf of the Entwicklungsgemeinschaft HTR at KFA Juelich, HRB Mannheim, INTERATOM Bensberg, HOBEG Hanau, and SIGRI Meitingen.

Turner, R.F.

1983-04-01

354

Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test  

NASA Astrophysics Data System (ADS)

One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

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

2014-10-01

355

U-GAS process for production of hydrogen from coal  

Microsoft Academic Search

Today, hydrogen is produced mainly from natural gas and petroleum fractions. Tomorrow, because reserves of natural gas and oil are declining while demand continues to increase, they cannot be considered available for long-term, large-scale production of hydrogen. Hydrogen obtained from coal is expected to be the lowest cost, large-scale source of hydrogen in the future. The U-GAS coal gasification process

R. J. Dihu; J. G. Patel

1982-01-01

356

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

Microsoft Academic Search

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 LiClKCl eutectic at 550 C and NaClKCl 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

Vladimir A Volkovich; Trevor R Griffiths; Robert C Thied

2003-01-01

357

Modification in Purex process using supported liquid membrane separation of cerium(III) via oxidation to cerium(IV)from fission products from nitrate medium by SLM  

Microsoft Academic Search

The separation of cerium from fission products during Purex process have been achieved successfully. Cerium(III) has been oxidized to Ce(IV) by 9M HNO3 and 2M NaBrO3, Ce(IV) was stable for a long time (6 months) and separated from Uranium(IV) completely. The separation factor was found to be 3. The effect of 30% TBPkerosene and different cations, Zr(IV) and Mo(VI) using

N. El-Said; N. Abdel Rahman; Emad H. Borai

2002-01-01

358

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

359

Computer model for production forecasting of oil and gas  

Microsoft Academic Search

In nature, gas generally is found in combination with oil. To make the situation even more complicated, the gas-oil ratio of the crude may vary from well to well, and from time to time within the same well. Without a computer program, the production manager must decide arbitrarily, from past experience and by trial and error, how much crude and

K. W. Hahn; S. A. Turaiki; A. S. Al-Mishari

1983-01-01

360

Relationship between hydrogen gas and butanol production by Clostridium saccharoperbutylacetonicum  

Microsoft Academic Search

Two simultaneous fermentations were performed at 26 degrees C with simultaneous inocula using Clostridium saccharoperbutylacetonicum. Fermentation 1 prevented the gas formed by the biomass from escaping the fermentor while 2 allowed the gas formed to escape. Fermentor 1 provided for the production of butanol, acetone, and ethanol, while when the H formed was allowed to escape with fermentor 2, neither

James D. Brosseau; Jwo-Yee Yan; K. Victor Lo

1986-01-01

361

21 CFR 173.350 - Combustion product gas.  

Code of Federal Regulations, 2011 CFR

...isooctane. The absorbance of the solution of combustion product gas shall not exceed that of the isooctane solvent at any wavelength in the specified range by more than one-third of the standard reference...

2011-04-01

362

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

363

Inert gas ups viscous oil production  

Microsoft Academic Search

One of the newest heavy oil-recovery techniques is that of cyclic injecting an inert gas of carbon dioxide and nitrogen in a producing well. This method has increased producing rates substantially in recent field tests. The experimental project is located in the Taylor Ina Field of Medina County in Southwest Texas. The wells previously pumped a few barrels per day,

Davison

1965-01-01

364

Tempest gas turbine extends EGT product line  

SciTech Connect

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

Chellini, R.

1995-07-01

365

Powering the World: Offshore Oil & Gas Production  

E-print Network

A ir, water, soil, food, biomas s Energy Solar,wi nd, geothermal, fossil, nucle ar,hydro Economy oil&gas supply Shale plays will also be producing an increasing part of global hydrocarbon supply Energy flow-based solutions (wind turbines, photovoltaics, and biofuels) will require most radical

Patzek, Tadeusz W.

366

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

SciTech Connect

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

Cho, Dong-Keun; Kim, Myung-Hyun [Kyung Hee University (Korea, Republic of)

2003-10-15

367

Mitigating Accidents In Oil And Gas Production Facilities  

Microsoft Academic Search

Integrated operations are increasingly used in oil and gas production facilities to improve yields, reduce costs and maximize\\u000a profits. They leverage information and communications technology (ICT) to facilitate collaboration between experts at widely\\u000a dispersed locations. This paper discusses the safety and security consequences of implementing integrated operations for oil\\u000a and gas production. It examines the increased accident risk arising from

Stig Johnsen

2009-01-01

368

Mitigating Accidents In Oil And Gas Production Facilities  

NASA Astrophysics Data System (ADS)

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

Johnsen, Stig

369

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

370

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

371

Advanced electron microscopic techniques applied to the characterization of irradiation effects and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment  

SciTech Connect

Preliminary electron microscopy of coated fuel particles from the AGR-1 experiment was conducted using characterization techniques such as scanning electron microscopy (SEM), transmission electron microscopy (TEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Microscopic quantification of fission-product precipitates was performed. Although numerous micro- and nano-sized precipitates observed in the coating layers during initial SEM characterization of the cross-sections, and in subsequent TEM diffraction patterns, were indexed as UPd{sub 2}Si{sub 2}, no Ag was conclusively found. Additionally, characterization of these precipitates highlighted the difficulty of measuring low concentrations of Ag in precipitates in the presence of significantly higher concentrations of Pd and U. The electron microscopy team followed a multi-directional and phased approach in the identification of fission products in irradiated TRISO fuel. The advanced electron microscopy techniques discussed in this paper, not only demonstrate the usefulness of the equipment (methods) as relevant research tools, but also provide relevant scientific results which increase the knowledge about TRISO fuel particles microstructure and fission products transport.

Rooyen, I.J. van; Lillo, T.M.; Trowbridge, T.L.; Madden, J.M. [Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Wu, Y.Q. [Boise State University, Boise, ID 83725-2090 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Goran, D. [Brucker Nano Gmbh, Berlin, 12489 (Germany)

2013-07-01

372

On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients  

NASA Astrophysics Data System (ADS)

The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved.

Walker, C. T.; Mogensen, M.

1987-07-01

373

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

Code of Federal Regulations, 2011 CFR

... false Rigid gas permeable contact lens care products. 886.5918 Section...5918 Rigid gas permeable contact lens care products. (a) Identification. A rigid gas permeable contact lens care product is a device...

2011-04-01

374

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

Code of Federal Regulations, 2010 CFR

... false Rigid gas permeable contact lens care products. 886.5918 Section...5918 Rigid gas permeable contact lens care products. (a) Identification. A rigid gas permeable contact lens care product is a device...

2010-04-01

375

Application of the Stretched Exponential Production Decline Model to Forecast Production in Shale Gas Reservoirs  

E-print Network

Production forecasting in shale (ultra-low permeability) gas reservoirs is of great interest due to the advent of multi-stage fracturing and horizontal drilling. The well renowned production forecasting model, Arps? Hyperbolic Decline Model...

Statton, James Cody

2012-07-16

376

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

USGS Publications Warehouse

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

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

2009-01-01

377

STABILIZING GLASS BONDED WASTE FORMS CONTAINING FISSION PRODUCTS SEPARATED FROM SPENT NUCLEAR FUEL  

SciTech Connect

A model has been developed to represent the stresses developed when a molten, glass-bonded brittle cylinder (used to store nuclear material) is cooled from high temperature to working temperature. Large diameter solid cylinders are formed by heating glass or glass-bonded mixtures (mixed with nuclear waste) to high temperature (915C). These cylinders must be cooled as the final step in preparing them for storage. Fast cooling time is desirable for production; however, if cooling is too fast, the cylinder can crack into many pieces. To demonstrate the capability of the model, cooling rate cracking data were obtained on small diameter (7.8 cm diameter) glass-only cylinders. The model and experimental data were combined to determine the critical cooling rate which separates the non-cracking stable glass region from the cracked, non-stable glass regime. Although the data have been obtained so far only on small glass-only cylinders, the data and model were used to extrapolate the critical-cooling rates for large diameter ceramic waste form (CWF) cylinders. The extrapolation estimates long term cooling requirements. While a 52-cm diameter cylinder (EBR-II-waste size) can be cooled to 100C in 70 hours without cracking, a 181.5-cm diameter cylinder (LWR waste size) requires 35 days to cool to 100C. These cooling times are long enough that verification of these estimates are required so additional experiments are planned on both glass only and CWF material.

Kenneth J. Bateman; Charles W. Solbrig

2008-07-01

378

NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS  

SciTech Connect

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

Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

2004-12-01

379

Production of low BTU gas from biomass  

E-print Network

) reported on gasification of coconut waste in the Philippines. Table 2 illustrates various gasifi- cation products that were obtained by different inves- tigators. Table 2. Comparison of Gasification Results from Various Sources (t)aterials Coconut...) reported on gasification of coconut waste in the Philippines. Table 2 illustrates various gasifi- cation products that were obtained by different inves- tigators. Table 2. Comparison of Gasification Results from Various Sources (t)aterials Coconut...

Lee, Yung N.

2012-06-07

380

Analysis of primary damage in silicon carbide under fusion and fission neutron spectra  

NASA Astrophysics Data System (ADS)

Irradiation parameters on primary damage states of SiC are evaluated and compared for the first wall of ITER under deuterium-deuterium (DD) and deuterium-tritium (DT) operation, the high temperature gas-cooled reactor (HTGR) and high flux isotope reactor (HFIR). With the same neutron fluence, the studied fusion spectra produce more damage and much higher gas production than the fission spectra. Due to comparable gas production and similar weighted primary recoil spectra, HFIR is considered suitable to simulate the neutron irradiation in an HTGR. In contrast to the significant differences between the weighted primary recoil spectra of the fission and the fusion spectra, the weighted secondary recoil spectra of HFIR and HTGR match those of DD and DT, indicating that displacement cascades by the fission and the fusion irradiation are similar when the damage distribution among damaged regions by secondary recoils is taken into account.

Guo, Daxi; Zang, Hang; Zhang, Peng; Xi, Jianqi; Li, Tao; Ma, Li; He, Chaohui

2014-12-01

381

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 simultaneously emits 14 MeV neutrons optimal to excite fast neutron-induced gamma-ray signals, such as from carbon and oxygen, and fission-like neutrons optimal to induce neutron capture gamma-ray signals, such as from hydrogen, nitrogen, and chlorine. Experiments were performed, which compare well to Monte Carlo simulations, showing that the uranium reflector enhances capture signals by up to a factor of 15 compared to the absence of a reflector.

Koltick, D.; McConchie, S.; Sword, E.

2008-04-01

382

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

SciTech Connect

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

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

1985-05-01

383

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

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

384

Production of Substitute Natural Gas from Coal  

SciTech Connect

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

Andrew Lucero

2009-01-31

385

Workshop of Arc By-Products in Gas Insulated Equipment  

NASA Astrophysics Data System (ADS)

A chemical data base was developed on the generation of arc by-products in gas insulated type equipment. Various aspects of arc induced decomposition of SF6, such as identification and quantitation of by-products, mechanisms of their formation and analytical methods used for their determination were reviewed. Users were made aware of the type of arc by-products encountered in SF6 insulated equipment, and to solicit their input with regard to the direction of any future work.

Tahiliani, V.; Vouros, P.

1980-12-01

386

Pumps, refracturing hike production from tight shale gas wells  

SciTech Connect

This paper reports that downhole pumps and refracturing are two ways to significantly improve production rates from the Antrim shale, a tight formation in the Michigan basin (U.S.) and the objective of a major natural gas play. Candidate wells for restimulation can be identified by pressure build-up tests and specifically productivity index-vs.-permeability plots based on these tests. The work in the Bagley East B4-10 well illustrates the possible production improvement.

Reeves, S.R. (Advanced Resources International Inc., Arlington, VA (United States)); Morrisson, W.K. (Nomeco Oil and Gas Co., Jackson, CO (United States)); Hill, D.G. (Gas Research Inst., Chicago, IL (United States))

1993-02-01

387

Future challenges for nuclear data research in fission (u)  

SciTech Connect

I describe some high priority research areas in nuclear fission, where applications in nuclear reactor technologies and in modeling criticality in general are demanding higher accuracies in our databases. We focus on fission cross sections, fission neutron spectra, and fission product data.

Chadwick, Mark B [Los Alamos National Laboratory

2010-01-01

388

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

SciTech Connect

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

NONE

1996-12-01

389

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

E-print Network

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

Torres-Verdín, Carlos

390

Shale Gas Production Theory and Case Analysis We researched the process of oil recovery and shale gas  

E-print Network

Shale Gas Production Theory and Case Analysis (Siemens) We researched the process of oil recovery and shale gas recovery and compare the difference between conventional and unconventional gas reservoir and recovery technologies. Then we did theoretical analysis on the shale gas production. According

Ge, Zigang

391

Photodissociation Dynamics of 2-BROMOETHYLNITRITE at 351 NM and C-C Bond Fission in the ? - Radical Product  

NASA Astrophysics Data System (ADS)

We used a crossed laser-molecular beam scattering experiment to investigate the primary photodissociation channels of bromoethylnitrite at 351 nm. Only the O-NO bond fission channel forming the ? -bromoethoxy radical and NO, no HBr photoelimination, was detected upon 351 nm photoexcitation,. The subsequent decomposition of the highly vibrational excited ? -bromoethoxy radical to formaldehyde + CH{_2}Br was also investigated.

Wang, Lei; Chhantyal-Pun, Rabi; Brynteson, Matt D.; Miller, Terry A.; Butler, Laurie J.

2013-06-01

392

Cascade heat recovery with coproduct gas production  

DOEpatents

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

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

1986-10-14

393

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

DOEpatents

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

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

2000-01-01

394

Fifty years with nuclear fission  

SciTech Connect

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

Behrens, J.W.; Carlson, A.D. (eds.) (National Institute of Standards and Technology, Gaithersburg, MD (United States))

1989-01-01

395

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

SciTech Connect

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

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

2011-06-02

396

Optimization and evaluation of mixed-bed chemisorbents for extracting fission and activation products from marine and fresh waters.  

PubMed

Chemically selective chemisorbents are needed to monitor natural and engineered waters for anthropogenic releases of stable and radioactive contaminants. Here, a number of individual and mixtures of chemisorbents were investigated for their ability to extract select fission and activation product elements from marine and coastal waters, including Co, Zr, Ru, Ag, Te, Sb, Ba, Cs, Ce, Eu, Pa, Np, and Th. Conventional manganese oxide and cyanoferrate sorbents, including commercially available Anfezh and potassium hexacyanocobalt(II) ferrate(II) (KCFC), were tested along with novel nano-structured surfaces (known as Self Assembled Monolayers on Mesoporous Supports or SAMMS) functionalized with a variety of moieties including thiol, diphosphonic acid (DiPhos-), methyl-3,4 hydroxypyridinone (HOPO-), and cyanoferrate. Extraction efficiencies were measured as a function of salinity, organic content, temperature, flow rate and sample size for both synthetic and natural fresh and saline waters under a range of environmentally relevant conditions. The effect of flow rate on extraction efficiency, from 1 to 70 mL min(-1), provided some insight on rate limitations of mechanisms affecting sorption processes. Optimized mixtures of sorbent-ligand chemistries afforded excellent retention of all target elements, except, Ba and Sb. Mixtures of tested chemisorbents, including MnO(2)/Anfezh and MnO(2)/KCFC/Thiol (1-3 mm)-SAMMS, extracted 8 of the 11 target elements studied to better than 80% efficiency, while a mixture of MnO(2)/Anfezh/Thiol (75-150 ?m)-SAMMS mixture was able to extract 7 of the 11 target elements to better than 90%. Results generated here indicate that flow rate should be less of a consideration for experimental design if sampling from fresh water containing variable amounts of DOM, rather than collecting samples from salt water environments. Relative to the capability of any single type of chemisorbent tested, optimized mixtures of several sorbents are able to increase the number of elements that can be efficiently and simultaneously extracted from natural waters. PMID:22093344

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

2011-12-01

397

Gas phase acetaldehyde production in a continuous bioreactor  

SciTech Connect

The gas phase continuous production of acetaldehyde was studied with particular emphasis on the development of biocatalyst (alcohol oxidase on solid phase support materials) for a fixed bed reactor. Based on the experimental results in a batch bioreactor, the biocatalysts were prepared by immobilization of alcohol oxidase on Amberlite IRA-400, packed into a column, and the continuous acetaldehyde production in the gas phase by alcohol oxidase was performed. The effects of the reaction temperature, flow rates of gaseous stream, and ethanol vapor concentration on the performance of the continuous bioreactor were investigated.

Hwang, Soon Ook (Northeastern Univ., Boston, MA (United States). Dept. of Chemical Engineering); Trantolo, D.J. (Northeastern Univ., Boston, MA (United States). Center for Biotechnology Engineering); Wise, D.L. (Northeastern Univ., Boston, MA (United States). Dept. of Chemical Engineering Northeastern Univ., Boston, MA (United States). Center for Biotechnology Engineering)

1993-08-20

398

Organic Sulfur Gas Production in Sulfidic Caves  

NASA Astrophysics Data System (ADS)

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

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

2001-12-01

399

Fifty years with nuclear fission  

SciTech Connect

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

Behrens, J.W.; Carlson, A.D. (eds.) (National Institute of Standards and Technology, Gaithersburg, MD (United States))

1989-01-01

400

Background radiation from fission pulses  

SciTech Connect

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

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

1988-05-01

401

Production of bio-synthetic natural gas in Canada.  

PubMed

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

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

2010-03-15

402

Microscopic description of complex nuclear decay: multimodal fission  

E-print Network

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

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

2009-06-23

403

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

NASA Astrophysics Data System (ADS)

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

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

2012-03-01

404

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

Microsoft Academic Search

Fission ⁹⁹Mo 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 μm copper foil instead of the uranium foil, wrapped in a 14μm nickel

R. Schrader; J. Klein; J. Medel; J. Marin; N. Salazar; M. Barrera; C. Albornoz; M. Chandia; X. Errazu; R. Becerra; G. Sylvester; J. C. Jimenez; E. Vargas

2008-01-01

405

Devonian shale gas production; Mechanisms and simple models  

SciTech Connect

This paper shows that, even without consideration of their special storage and flow properties, Devonian shales are special cases of dual porosity. The authors show that wile neglecting these properties in the short term is appropriate, such neglect in the long term will result in an under-estimation of shale gas production.

Carlson, E.S. (Univ. of Alabama (US)); Mercer, J.C. (Dept. of Energy (US))

1991-04-01

406

THE NUMBER OF NEUTRONS EMITTED BY U²³⁵ IN THE REGION OF SYMMETRICAL FISSION  

Microsoft Academic Search

The change in the number of neutrons emitted in the process of fission ; was investigated for various ratios of the fission product masses. A double ; ionization chamber having grids to measure the mass distribution of fission ; products was used to detect the fission products obtained by the fis-sion of U\\/; sup 235\\/ by thermal neutrons. A 4

V. F. Apalin; Yu. N. Gritsyuk; I. E. Kutikov; V. I. Lebedev; L. A. Mikaelyan

1962-01-01

407

Alaska North Slope regional gas hydrate production modeling forecasts  

USGS Publications Warehouse

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

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

2011-01-01

408

Dynamical Aspects of Nuclear Fission  

NASA Astrophysics Data System (ADS)

Fission dynamics. Dependence of scission-neutron yield on light-fragment mass for [symbol]=1/2 [et al.]. Dynamics of capture quasifission and fusion-fission competition / L. Stuttg ... [et al.] -- Fission-fission. The processes of fusion-fission and quasi-fission of superheavy nuclei / M. G. Itkis ... [et al.]. Fission and quasifission in the reactions [symbol]Ca+[symbol]Pb and [symbol]Ni+[symbol]W / G. N. Knyazheva ... [et al.]. Mass-energy characteristics of reactions [symbol]Fe+[symbol][symbol][symbol]266Hs and [symbol]Mg+[symbol]Cm[symbol][symbol]Hs at Coulomb barrier / L. Krupa ... [et al.]. Fusion of heavy ions at extreme sub-barrier energies / ?. Mi?icu and H. Esbensen. Fusion and fission dynamics of heavy nuclear system / V. Zagrebaev and W. Greiner. Time-dependent potential energy for fusion and fission processes / A. V. Karpov ... [et al.] -- Superheavy elements. Advances in the understanding of structure and production mechanisms for superheavy elements / W. Greiner and V. Zagrebaev. Fission barriers of heaviest nuclei / A. Sobiczewski ... [et al.]. Possibility of synthesizing doubly magic superheavy nuclei / Y Aritomo ... [et al.]. Synthesis of superheavy nuclei in [symbol]Ca-induced reactions / V. K. Utyonkov ... [et al.] -- Fragmentation. Production of neutron-rich nuclei in the nucleus-nucleus collisions around the Fermi energy / M. Veselsk. Signals of enlarged core in [symbol]Al / Y. G. Ma ... [et al.] -- Exotic modes. New insight into the fission process from experiments with relativistic heavy-ion beams / K.-H. Schmidt ... [et al.]. New results for the intensity of bimodal fission in binary and ternary spontaneous fission of [symbol]Cf / C. Goodin ... [et al.]. Rare fission modes: study of multi-cluster decays of actinide nuclei / D. V. Kamanin ... [et al.]. Energy distribution of ternary [symbol]-particles in [symbol]Cf(sf) / M. Mutterer ... [et al.]. Preliminary results of experiment aimed at searching for collinear cluster tripartition of [symbol]Pu / Y. V. Pyatkov. Comparative study of the ternary particle emission in [symbol] and [symbol]Cm(SF) / S. Vermote ... [et al.] -- Structure of fission fragments and neurton rich nuclei / manifestation of average y-ray multiplicity in the fission modes of [symbol]Cf(sf) and the proton-induced fission of [symbol]Pa, [symbol]Np, and [symbol]Am / M. Bereov ... [et al.]. Yields of correlated fragment pairs and average gamma-ray multiplicities and energies in [symbol]Pb([symbol]O, f) / A. Bogachev ... [et al.]. Recent experiments at gammasphere intended to the study of [symbol]Cf spontaneous fission / A. V. Daniel ... [et al.]. Nuclear structure studies of microsecond isomers near A =100 / J. Genevey ... [et al.]. Covariant density functional theory: isospin properties of nuclei far from stability / G. A. Lalazisis. Relativistic mean-field description of light nuclei / J. Leja and . Gmuca. Energy nucleon spectra from reactions at intermediate energies / O. Grudzevich ... [et al.] -- Developments in experimental techniques. Analysis, processing and visualization of multidimensional data using DaqProVis system / M. Morh? ... [et al.].

Kliman, J.; Itkis, M. G.; Gmuca, .

2008-11-01

409

Separation of Long-Lived Fission Products Tc-99 and I-129 from Synthetic Effluents by Crown Ethers  

SciTech Connect

To minimize significantly the radio-toxic inventory of nuclear geological repositories to come as well as to reduce the potential of radionuclides migration and to minimize long-term exposure, the concept of partitioning and transmutation (P/T) of nuclear waste is currently discussed. Transmutation offers the possibility to convert radio-toxic radionuclides with long half-lives into radionuclides of shorter half-lives, less toxic isotopes, or even into stable isotopes. Besides the most prominent isotopes of neptunium, plutonium, americium, and curium, the long-lived fission products Tc-99 and I-129 (half-lives of 2.13 x 10{sup 5} years, and 1.57 x 10{sup 7} years, respectively) are promising candidates for transmutation in order to prevent their migration from a nuclear repository. Partitioning and transmutation of the most radio-toxic radionuclides will not only minimize the nuclear waste load but most importantly will significantly reduce the long-term radio-toxic hazard of nuclear waste repositories to come. Prior to the deployment of partitioning and transmutation, selective extraction techniques are required to separate the radionuclides of concern. Since the discovery of crown ethers by C. Pedersen, various applications of crown ethers have drawn much attention. Although liquid-liquid extraction of alkali and alkali earth metals by crown ethers has been extensively studied, little data is available on the extraction of Tc-99 and I-129 by crown ethers. The methods developed herein for the specific extraction of Tc-99 and I-129 provide recommendations in support of their selectively extraction from liquid radioactive waste streams, mainly ILW. We report data on the solvent extraction of Tc-99 and I-129 from synthetic effluents by six crown ethers of varying cavity dimensions and derivatization. To satisfy the needs of new extractant systems we are demonstrating that crown ether (CE) based systems have the potential to serve as selective extractants for the separation of these long lived radionuclides from high level nuclear waste (HLW), intermediate level nuclear waste (ILW), and low level nuclear waste (LLW) streams. The experimental results show that dibenzo-18-crown-6 (DB 18C6) is highly selective towards Tc-99, and dicyclohexano-18-crown-6 (DC18C6) is highly selective towards I-129. The nature of the diluent was examined and was shown to be the most influential variable in controlling the extraction coefficients of Tc-99 and I-129. Therefore the addition of polar diluent acetone to non-polar diluent toluene enhanced the distribution coefficient of Tc-99 (DTc) was by a factor of 30. For I-129, the best extraction yield was obtained after introducing tetrachloroethane. Through the process, by a single extraction step, 85 % to 95 % of Tc-99 was extracted from synthetic effluents, while 84 % to 88 % of I-129 was extracted from different acidic media. The extraction by crown ether is a fairly rapid process and the total preparation time of the chemical separation takes about 20 minutes for a batch of eight samples. (authors)

Paviet-Hartmann, P. [Framatome-ANP Inc. an AREVA and Siemens Company, 128 S. Tryon, FC 12A, Charlotte, NC 28202 (United States); Hartmann, T. [University of Nevada - Las Vegas, Harry Reid Center, 4505 Maryland Parkway, Box 454009, Las Vegas, NV 89154 (United States)

2006-07-01

410

Gas plant economic optimization is more than meeting product specification  

SciTech Connect

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

Berkowitz, P.N.; Colwell, L.W. [Continental Controls, Inc., Houston, TX (United States); Gamez, J.P. [Gas Research Inst., Chicago, IL (United States)

1996-12-31

411

NOBLE GAS PRODUCTION FROM MERCURY SPALLATION AT SNS  

SciTech Connect

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

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

2013-01-01

412

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

NASA Technical Reports Server (NTRS)

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

Caraccio, Anne J.; Hintze, Paul E.

2013-01-01

413

Value-Added Products from Remote Natural Gas  

SciTech Connect

In Wyoming and throughout the United States, there are natural gas fields that are not producing because of their remoteness from gas pipelines. Some of these fields are ideal candidates for a cogeneration scheme where components suitable for chemical feedstock or direct use, such as propane and butane, are separated. Resulting low- to medium-Btu gas is fired in a gas turbine system to provide power for the separation plant. Excess power is sold to the utility, making the integrated plant a true cogeneration facility. This project seeks to identify the appropriate technologies for various subsystems of an integrated plant to recover value-added products from wet gas and/or retrograde condensate reservoirs. Various vendors and equipment manufacturers will be contacted and a data base consisting of feedstock constraints and output specifications for various subsystems and components will be developed. Based on vendor specifications, gas reservoirs suited for value-added product recovery will be identified. A candidate reservoir will then be selected, and an optimum plant layout will be developed. A facility will then be constructed and operated. The project consists of eight subtasks: Compilation of Reservoir Data; Review of Treatment and Conditioning Technologies; Review of Product Recovery and Separation Technologies; Development of Power Generation System; Integrated Plant Design for Candidate Field; System Fabrication; System Operation and Monitoring; and Economic Evaluation and Reporting. The first five tasks have been completed and the sixth is nearly complete. Systems Operations and Monitoring will start next year. The Economic Evaluation and Reporting task will be a continuous effort for the entire project. The reservoir selected for the initial demonstration of the process is the Burnt Wagon Field, Natrona County, Wyoming. The field is in a remote location with no electric power to the area and no gas transmission line. The design for the gas processing train to produce the liquefied gas products includes three gas compressors, a cryogenic separation unit, and a natural gas powered generator. Based on the equipment specifications, air quality permits for the well field and the gas processing unit were developed and the permits were issued by the Wyoming Department of Environmental Quality. Also, to make state and federal reporting easier, three of the four leases that made up the Burnt Wagon were combined. All major equipment has been installed and individual component operability is being conducted. During the next project year, operability testing and the shakedown of the entire system will be completed. Once shakedown is complete, the system will be turned over to the cosponsor for day-to-day operations. During operations, data will be collected through remote linkage to the data acquisition system or analysis of the system performance to develop an economic evaluation of the process.

Lyle A. Johnson

2002-03-15

414

Advanced Space Fission Propulsion Systems  

NASA Technical Reports Server (NTRS)

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

Houts, Michael G.; Borowski, Stanley K.

2010-01-01

415

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

E-print Network

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

Diecker, Jane T

2005-01-01

416

Measurement\\/Evaluation Techniques and Nuclear Data Associated with Fission of 239Pu by Fission Spectrum Neutrons  

Microsoft Academic Search

This Panel was chartered to review and assess new evaluations of work on fission product data, as well as the evaluation process used by the two U.S. nuclear weapons physics laboratories. The work focuses on fission product yields resulting from fission spectrum neutrons incident on plutonium, and includes data from measurements that had not been previously published as well as

P Baisden; E Bauge; J Ferguson; D Gilliam; T Granier; R Jeanloz; C McMillan; D Robertson; P Thompson; C Verdon; C Wilkerson; P Young

2010-01-01

417

Accounting for product financing arrangements by oil and gas producers  

SciTech Connect

The Financial Accounting Standards Board (FASB) has developed the Statement of Financial2 Accounting Standards (SFAS) Nos. 47 and 49 to help practitioners in accounting for and disclosing product financing arrangements in the oil and gas industry. SFAS No. 47 is a disclosure document only, while SFAS No. 49 specifies the accounting treatment for certain arrangements. The authors describe and give examples to show how practitioners can implement the substantive provisions of the documents.

Munter, P.; Ratcliffe, T.A.

1983-03-01

418

Continuous ethanol production in the gas-lift tower fermenter  

Microsoft Academic Search

A highly flocculent strain of Saccharomyces uvarum was used to convert glucose to ethanol and CO2 in a single stage, continuous, gas-lift tower fermenter. Satisfactory operation was maintained in prolonged runs with yeast concentrations in excess of 100 g\\/L (d.w.) and hydraulic retention times less than 0.4 h. Maximum ethanol concentration and productivity were 88 g\\/L and 44.5 g\\/Lh respectively.

D. Martin Comberbach; John D. Bu'Lock

1984-01-01

419

Relationship between hydrogen gas and butanol production by Clostridium saccharoperbutylacetonicum  

SciTech Connect

Two simultaneous fermentations were performed at 26 degrees C with simultaneous inocula using Clostridium saccharoperbutylacetonicum. Fermentation 1 prevented the gas formed by the biomass from escaping the fermentor while 2 allowed the gas formed to escape. Fermentor 1 provided for the production of butanol, acetone, and ethanol, while when the H/sub 2/ formed was allowed to escape with fermentor 2, neither butanol nor acetone were produced. Ethanol was also formed in both fermentors and began along with the initial growth of biomass and continued until the fermentations were complete. Butanol and acetone production began after biomass growth had reached a maximum and began to subside. The butanol-acetone-ethanol millimolar yields and ratios were 38:1:14 respectively. The fermentor 2 results show that a yield of 2.1 l H/sub 2/, 93 or 370 mmol H/sub 2//mol glucose, was formed only during the growing stage of growth; neither butanol nor acetone were produced; ethanol was formed throughout the fermentation, reaching a yield of 15.2 mmolar. It appears that hydrogen gas is required for butanol production during the resting stage of growth. 16 references.

Brosseau, J.D.; Yan, J.Y.; Lo, K.V.

1986-03-01

420

Simulation of natural gas production from submarine gas hydrate deposits combined with carbon dioxide storage  

NASA Astrophysics Data System (ADS)

The recovery of methane from gas hydrate layers that have been detected in several submarine sediments and permafrost regions around the world so far is considered to be a promising measure to overcome future shortages in natural gas as fuel or raw material for chemical syntheses. Being aware that natural gas resources that can be exploited with conventional technologies are limited, research is going on to open up new sources and develop technologies to produce methane and other energy carriers. Thus various research programs have started since the early 1990s in Japan, USA, Canada, South Korea, India, China and Germany to investigate hydrate deposits and develop technologies to destabilize the hydrates and obtain the pure gas. In recent years, intensive research has focussed on the capture and storage of carbon dioxide from combustion processes to reduce climate change. While different natural or manmade reservoirs like deep aquifers, exhausted oil and gas deposits or other geological formations are considered to store gaseous or liquid carbon dioxide, the storage of carbon dioxide as hydrate in former methane hydrate fields is another promising alternative. Due to beneficial stability conditions, methane recovery may be well combined with CO2 storage in form of hydrates. This has been shown in several laboratory tests and simulations - technical field tests are still in preparation. Within the scope of the German research project SUGAR, different technological approaches are evaluated and compared by means of dynamic system simulations and analysis. Detailed mathematical models for the most relevant chemical and physical effects are developed. The basic mechanisms of gas hydrate formation/dissociation and heat and mass transport in porous media are considered and implemented into simulation programs like CMG STARS and COMSOL Multiphysics. New simulations based on field data have been carried out. The studies focus on the evaluation of the gas production potential from turbidites and their ability for carbon dioxide storage. The effects occurring during gas production and CO2 storage within a hydrate deposit are identified and described for various scenarios. The behaviour of relevant process parameters such as pressure, temperature and phase saturations is discussed and compared for different production strategies: depressurization, CO2 injection after depressurization and simultaneous methane production and CO2 injection.

Janicki, Georg; Schlter, Stefan; Hennig, Torsten; Deerberg, Grge

2013-04-01