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1

New Results on Helium and Tritium Gas Production From Ternary Fission  

SciTech Connect

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

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

2005-05-24

2

Fission product range effects on HEU fissile gas monitoring for UF{sub 6} gas  

SciTech Connect

The amount of {sup 235}U in UF{sub 6} flowing in a pipe can be monitored by counting gamma rays emitted from fission fragments carried along by the flowing gas. Neutron sources are mounted in an annular sleeve that is filled with moderator material and surrounds the pipe. This provides a source of thermal neutrons to produce the fission fragments. Those fragments that remain in the gas stream following fission are carried past a gamma detector. A typical fragment will be quite unstable, giving up energy as it decays to a more stable isotope with a significant amount of this energy being emitted in the form of gamma rays. A given fragment can emit several gamma rays over its lifetime. The gamma ray emission activity level of a distribution of fission fragments decreases with time. The monitoring system software uses models of these processes to interpret the gamma radiation counting data measured by the gamma detectors.

Munro, J.K. Jr.; Valentine, T.E.; Perez, R.B. [and others

1997-09-01

3

Fission gas detection system  

DOEpatents

A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

Colburn, Richard P. (Pasco, WA)

1985-01-01

4

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

SciTech Connect

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

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

2010-10-01

5

Fission Product 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/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

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

2008-09-01

6

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

SciTech Connect

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

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

2010-10-01

7

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

NASA Technical Reports Server (NTRS)

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

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

1974-01-01

8

Fission-product retention in HTGR fuels  

SciTech Connect

Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.

Homan, F.J.; Kania, M.J.; Tiegs, T.N.

1982-01-01

9

Fission product solvent extraction  

SciTech Connect

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

Moyer, B.A.; Bonnesen, P.V.; Sachleben, R.A. [and others

1998-02-01

10

THE RETURN OF ESCAPED FISSION PRODUCT GASES TO UOâ  

Microsoft Academic Search

Preliminary experimental results appear consistent with the hypothesis ; of J. A. Davies by which fission fragments passing through a space filled with ; escaped fission product gases knock on a certain number of the gas atoms with ; sufficient energy, 2 kev or more, to penetrate through solid surfaces and become ; trapped. The number of gas atoms knocked

1960-01-01

11

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

12

FFTF fission gas monitor computer system  

Microsoft Academic Search

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled 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 control room

Hubbard

1987-01-01

13

Measurement of Fission Product Yields from Fast-Neutron Fission  

NASA Astrophysics Data System (ADS)

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

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

2014-09-01

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

Modeling Fission Product Sorption in Graphite Structures  

SciTech Connect

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).

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

2013-04-08

16

Recovery and use of fission product noble metals  

SciTech Connect

Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value. (DLC)

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

1980-06-01

17

AMS measurements of fission products at CIAE  

NASA Astrophysics Data System (ADS)

Fission products are present in special nuclear materials as contaminants remaining from isotope separation or reprocessing, or through ingrowth due to spontaneous and neutron induced fission. The long half-lived fission products (LLFPs) are among the most dangerous radionuclides to the environment. Ultra-high-sensitivity measurement of LLFPs in rocks or soil samples from the fission environment would provide very important information for nuclear safety inspection. The Beijing HI-13-AMS facility with a high terminal voltage of 13 MV is suitable for measuring LLFPs, especially for heavy fission products such as 79Se, 93Zr, 99Tc, 107Pd, 121mSn, 126Sn, 129I and 151Sm. In this paper some new methods developed for AMS measurement of 79Se, 93Zr, 99Tc, 121mSn, 126Sn, 129I and 151Sm are presented. Major features of these methods will be introduced, including the preparation of samples, the selection of target material and the molecular ions extracted from the material in the ion source, as well as the identification and detection of the nuclides to be determined.

Shen, Hongtao; Jiang, Shan; He, Ming; Dong, Kejun; Ouyang, Yinggen; Li, Zhenyu; Guan, Yongjing; Yin, Xinyi; Peng, Bo; Zhou, Duo; Yuan, Jian; Wu, Shaoyong

2013-01-01

18

Fission gas bubble behaviour in uranium dioxide  

NASA Astrophysics Data System (ADS)

A theoretical model developed to study the gas bubble evolution in nuclear materials has been used to analyze experiments on uranium dioxide irradiated to low burnups (0.1 and 0.4 at%), in which fission gas bubble size distributions were measured following out-of-pile isothermal anneals. Following irradiation, the UO 2 was annealed for 1 or 6 h each at temperatures between 1303 and 1973 K and then thinned for transmission electron microscopy observation of the bubble size distributions. The model is based on the assumption, that the coalescence of the moving bubbles is the main mechanism defining gas porosity development under these conditions. The gas bubbles are assumed to be in equilibrium and their motion is assumed to be caused by random migration. The calculations show that the observed bubble size distributions may be reproduced on the base of the bubble growth mechanisms considered. The joint action of bubble surface and volume diffusion mechanisms can explain both the general nature of the experimental distributions and their perculiarities, in particularly the bimodal bubble size distribution observed after annealing at 1673 K. The choice of appropriate values as input parameters into the model is discussed.

Chkuaseli, V. F.; Matzke, Hj.

19

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

20

Fission gas release restrictor for breached fuel rod  

DOEpatents

In the event of a breach in the cladding of a rod in an operating liquid metal fast breeder reactor, the rapid release of high-pressure gas from the fission gas plenum may result in a gas blanketing of the breached rod and rods adjacent thereto which impairs the heat transfer to the liquid metal coolant. In order to control the release rate of fission gas in the event of a breached rod, the substantial portion of the conventional fission gas plenum is formed as a gas bottle means which includes a gas pervious means in a small portion thereof. During normal reactor operation, as the fission gas pressure gradually increases, the gas pressure interiorly of and exteriorly of the gas bottle means equalizes. In the event of a breach in the cladding, the gas pervious means in the gas bottle means constitutes a sufficient restriction to the rapid flow of gas therethrough that under maximum design pressure differential conditions, the fission gas flow through the breach will not significantly reduce the heat transfer from the affected rod and adjacent rods to the liquid metal heat transfer fluid flowing therebetween.

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

1986-01-01

21

Fission product yields in the fast-neutron fission of 238 U  

Microsoft Academic Search

The fission yields of 38 fission products in the fast-neutron induced fission of238U have been determined using a rapid, multiscaling gamma-ray spectroscopic method. To obtain absolute yields for fission products having half-lives ranging from 32 s to 40 d, a total of 56 multi-scaling gamma-ray spectra were collected using various irradiation and cooling periods. Gamma-rays and photopeak areas of interest

Chien Chung; Ming-Yung Woo

1987-01-01

22

PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA  

E-print Network

was slow. PASSAGE OF FISSION PRODUCTS THROUGH THE SKIN OF TUNA In relation to "fallout" from nuclear -bomb the penetration of radioactive strontium, cesium, and ruthenium common products of nuclear fission, through

23

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

24

Sensitivity analysis of the fission gas behavior model in BISON.  

SciTech Connect

This report summarizes the result of a NEAMS project focused on sensitivity analysis of a new model for the fission gas behavior (release and swelling) in the BISON fuel performance code of Idaho National Laboratory. Using the new model in BISON, the sensitivity of the calculated fission gas release and swelling to the involved parameters and the associated uncertainties is investigated. The study results in a quantitative assessment of the role of intrinsic uncertainties in the analysis of fission gas behavior in nuclear fuel.

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

2013-05-01

25

Fractal Model of Fission Product Release in Nuclear Fuel  

NASA Astrophysics Data System (ADS)

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

Stankunas, Gediminas

2012-09-01

26

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

27

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

NASA Technical Reports Server (NTRS)

A sweep gas facility has been used to measure the release rates of radioactive fission gases from small UN specimens irradiated to 8-percent burnup at high fission-rate densities. The measured release rates have been correlated with an equation whose terms correspond to direct recoil release, fission-enhanced diffusion, and atomic diffusion (a function of temperature). Release rates were found to increase linearly with burnups between 1.5 and 8 percent. Pore migration was observed after operation at 1550 K to over 6 percent burnup.

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

1973-01-01

28

Modeling of Fission Gas Release in UO2  

SciTech Connect

A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

MH Krohn

2006-01-23

29

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

30

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

31

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

32

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

SciTech Connect

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

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

2006-10-01

33

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

34

Fission products stability in uranium dioxide  

NASA Astrophysics Data System (ADS)

Fission product stability in nuclear fuels is investigated using density functional theory (DFT). In particular, incorporation and solution energies of He, Kr, Xe, I, Te, Ru, Sr and Ce in pre-existing trap sites of UO 2 (vacancies, interstitials, U-O divacancy, and Schottky trio defects) are calculated using the projector-augmented-wave method as implemented in the Vienna ab initio simulation package. Correlation effects are taken into account within the DFT+U approach. The stability of many binary and ternary compounds in comparison to soluted atoms is also explored. Finally the involvement of FP in the formation of metallic and oxide precipitates in oxide fuels is discussed in the light of experimental results.

Brillant, G.; Gupta, F.; Pasturel, A.

2011-05-01

35

(Fuel, fission product, and graphite technology)  

SciTech Connect

Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.

Stansfield, O.M.

1990-07-25

36

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 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

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

2013-03-01

37

Bulk and surface controlled diffusion of fission gas atoms  

SciTech Connect

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

Andersson, Anders D. [Los Alamos National Laboratory

2012-08-09

38

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

NASA Astrophysics Data System (ADS)

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

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

2005-12-01

39

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

40

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

Microsoft Academic Search

A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7MeV neutrons that produce fission events and their ?-delayed neutron emission or ?-delayed high-energy ? radiation between beam pulses provide the detection signature. Fission product ?-delayed ?-rays above 3MeV are

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

2005-01-01

41

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

Microsoft Academic Search

A concept for detecting the presence of special nuclear material (²³U or ²³Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their -delayed neutron emission or -delayed high-energy -radiation between beam pulses provide the detection signature. Fission product -delayed -rays above 3 MeV

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

2004-01-01

42

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

SciTech Connect

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

Tsekhanovich, I.; Varapai, N.; Rubchenya, V.; Rochman, D.; Simpson, G.S.; Sokolov, V.; Fioni, G.; Al Mahamid, Ilham [Institut Laue-Langevin, 38042 Grenoble (France); Petersburg Nuclear Physics Institute, 188350 Gatchina (Russian Federation); Commissariat a l'Energie Atomique, Siege, 75752 Paris Cedex 15 (France); Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States)

2004-10-01

43

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

J M Harp; P D Demkowicz; S A Ploger

2012-10-01

44

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

45

Determination of 140La fission product interference factor for INAA  

NASA Astrophysics Data System (ADS)

Instrumental Neutron Activation Analysis (INAA) is a technique widely used to determine the concentration of several elements in several kinds of matrices. However if the sample of interest has higher relative uranium concentration the obtained results can be interfered by the uranium fission products. One of these cases that is affected by interference due to U fission is the 140La , because this radioisotope used in INAA for the determination of concentration the La is also produced by the -? of 140Ba , an uranium fission product. The 140La interference factor was studied in this work and a factor to describe its time dependence was obtained.

Ribeiro, Iberê S., Jr.; Genezini, Frederico A.; Saiki, Mitiko; Zahn, Guilherme S.

2014-11-01

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

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

48

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

NASA Astrophysics Data System (ADS)

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

Padgett, Stephen; Wang, Tzu-Fang

2015-01-01

49

Fission product distribution in oxide fuels (LWBR Development Program)  

Microsoft Academic Search

Radial electron probe traverses were made on samples of ThO and ThO-20 w\\/o UO after approximately 15 and 21 x 10²° fissions\\/cm³, respectively. Concentration profiles of ²³³U and fission products in the thoria sample, which had not developed columnar grains, varied approximately linearly with the pellet radius. The slope of the profile varied with the volatility of the nuclide being

Berman

1976-01-01

50

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

51

Comparison of Fission Product Yields and Their Impact  

SciTech Connect

This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiological transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.

S. Harrison

2006-02-01

52

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

SciTech Connect

A concept for detecting the presence of special nuclear material ({sup 235}U or {sup 239}Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their {beta}-delayed neutron emission or {beta}-delayed high-energy {gamma}-radiation between beam pulses provide the detection signature. Fission product {beta}-delayed {gamma}-rays above 3 MeV are nearly ten times more abundant than {beta}-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified. An important goal in the US is the detection of nuclear weapons or special nuclear material (SNM) concealed in intermodal cargo containers. This must be done with high detection probability, low false alarm rates, and without impeding commerce, i.e. about one minute for an inspection. The concept for inspection has been described before and its components are now being evaluated. While normal radiations emitted from plutonium may allow its detection, the majority of {sup 235}U {gamma} ray emission is at 186 keV, is readily attenuated by cargo, and thus not a reliable detection signature for passive detection. Delayed neutron detection following a neutron or photon beam pulse has been used successfully to detect lightly or unshielded SNM targets. While delayed neutrons can be easily distinguished from beam neutrons they have relatively low yield in fission, approximately 0.008 per fission in {sup 239}Pu and 0.017 per fission in {sup 235}U, and are rapidly attenuated in hydrogenous materials making that technique unreliable when challenged by thick hydrogenous cargo overburden. They propose detection of {beta}-delayed high-energy {gamma} radiation as a more robust signature characteristic of SNM.

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

2004-09-30

53

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

NASA Astrophysics Data System (ADS)

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

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

2010-12-01

54

Transient fission gas behavior in uranium nitride fuel under proposed space applications  

Microsoft Academic Search

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

Daniel L. Deforest

1991-01-01

55

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

Microsoft Academic Search

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

Deforest

1991-01-01

56

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

SciTech Connect

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

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

2011-07-01

57

(Fission product transport experiments (HFR-B1))  

SciTech Connect

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

Myers, B.F.

1989-12-05

58

THE BEHAVIOR OF FISSION PRODUCTS IN MOLTEN FLUORIDE REACTOR FUELS  

Microsoft Academic Search

Observations are reported on the behavior of several fission product ; elements in molten NaF-ZrFâ-UFâ fuels, irradiated in capsule ; experiments, forced-con vection in-pile loop experiments, and in the Aircraft ; Reactor Experiment (ARE). The rare gases were observed to escape readily from ; the fuels in dynamic tests, although in static tests the rate of escape is very ;

M. T. Robinson; W. A. Jr. Brooksbank; S. A. Reynolds; H. W. Wright; T. H. Handley

1958-01-01

59

Fission product release from highly irradiated LWR fuel  

SciTech Connect

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

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

1980-02-01

60

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 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

Dawn Scates

2010-10-01

61

Recent MELCOR and VICTORIA Fission Product Research at the NRC  

SciTech Connect

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

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

1999-01-21

62

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

SciTech Connect

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

Blaise Collin

2014-09-01

63

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

E-print Network

November, 1967 Riso Report No. 170 Investigations on the Plant Uptake of Fission Products from Figures 23 #12;Dl AMI/ DLAINK #12;- 3 - INTRODUCTION The uptake of fission products by plants constitutes - plant relationships of different fission products have shown that the long-lived isotopes of stron- tium

64

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

NASA Astrophysics Data System (ADS)

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

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

2015-01-01

65

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

NASA Astrophysics Data System (ADS)

Renewed interest in fast nuclear reactors is creating a need for better understanding of fission gas bubble behavior in non-oxide fuels to support very long fuel lifetimes. Collisions between fission fragments and their subsequent cascades can knock fission gas atoms out of bubbles and back into the fuel lattice. We showed that these collisions can be treated as using the so-called 'homogenous' atom-by-atom re-solution theory and calculated using the Binary Collision Approximation code 3DOT. The calculations showed that there is a decrease in the re-solution parameter as bubble radius increases until about 50 nm, at which the re-solution parameter stays nearly constant. Furthermore, our model shows ion cascades created in the fuel result in many more implanted fission gas atoms than collisions directly with fission fragments. This calculated re-solution parameter can be used to find a re-solution rate for future bubble simulations.

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

2015-02-01

66

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

SciTech Connect

Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

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

1983-01-01

67

Fission product precipitates in irradiated uranium carbonitride fuel  

NASA Astrophysics Data System (ADS)

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

Kleykamp, H.

2002-02-01

68

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

DOEpatents

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

Herrmann, Steven Douglas

2014-05-27

69

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

70

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

71

Fission product ion exchange between zeolite and a molten salt  

NASA Astrophysics Data System (ADS)

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

Gougar, Mary Lou D.

72

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

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

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR  

Microsoft Academic Search

The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this

D. D. Jensen; M. J. Haire; A. Ballagny

1978-01-01

75

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

76

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

Microsoft Academic Search

A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7 MeV neutrons that produce fission events and their beta-delayed neutron emission or beta-delayed high-energy gamma radiation between beam pulses provide the detection signature. Fission product beta-delayed gamma-rays above 3

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

2005-01-01

77

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

SciTech Connect

A unique set of fission product gamma spectra was collected at short times (4 minutes to 1 week) on various fissionable materials. Gamma spectra were collected from the neutron-induced fission of uranium, neptunium, and plutonium isotopes at thermal, epithermal, fission spectrum, and 14-MeV neutron energies. This report describes the experimental methods used to produce and collect the gamma data, defines the experimental parameters for each method, and demonstrates the consistency of the measurements.

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

2011-09-29

78

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

79

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

SciTech Connect

The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures.

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

1983-10-01

80

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

SciTech Connect

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

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

2005-05-24

81

A proposed standard on medical isotope production in fission reactors  

SciTech Connect

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

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

2006-07-01

82

FISSION PRODUCT RADIOACTIVITY IN THE AIR ALONG THE 80th MERIDIAN, JANUARY JUNE 1957  

Microsoft Academic Search

Measurements of gross fission product radioactivity in the air at a ; number of sites along the 8Oth meridian (west) are reported for the period ; January to June 1957. The concentration of long-lived radioactive products ; (primarily fission products) in the air continues to reiqain considerably higher ; in the Northern than in the Southern Hemisphere. Nowhere has it

L. B. Jr. Lockhart; R. A. Baus; I. H. Jr. Bilifford

1957-01-01

83

Biological removal of cationic fission products from nuclear wastewater.  

PubMed

Nuclear energy is becoming a preferred energy source amidst rising concerns over the impacts of fossil fuel based energy on global warming and climate change. However, the radioactive waste generated during nuclear power generation contains harmful long-lived fission products such as strontium (Sr). In this study, cationic strontium uptake from solution by microbial cultures obtained from mine wastewater is evaluated. A high strontium removal capacity (q(max)) with maximum loading of 444 mg/g biomass was achieved by a mixed sulphate reducing bacteria (SRB) culture. Sr removal in SRB was facilitated by cell surface based electrostatic interactions with the formation of weak ionic bonds, as 68% of the adsorbed Sr(2+) was easily desorbed from the biomass in an ion exchange reaction with MgCl?. To a lesser extent, precipitation reactions were also found to account for the removal of Sr from aqueous solution as about 3% of the sorbed Sr was precipitated due to the presence of chemical ligands while the remainder occurred as an immobile fraction. Further analysis of the Sr-loaded SRB biomass by scanning electron microscopy (SEM) coupled to energy dispersive X-ray (EDX) confirmed extracellular Sr(2+) precipitation as a result of chemical interaction. In summary, the obtained results demonstrate the prospects of using biological technologies for the remediation of industrial wastewaters contaminated by fission products. PMID:21245563

Ngwenya, N; Chirwa, E M N

2011-01-01

84

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

85

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

86

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

PubMed

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

Matsson, I; Grapengiesser, B; Andersson, B

2007-01-01

87

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

88

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

DOEpatents

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

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

1987-10-26

89

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

90

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

SciTech Connect

Fission yields are especially well characterized for long-lived fission products. Modeling techniques incorporate numerous assumptions and can be used to deduce information about the distribution of short-lived fission products. This work is an attempt to gather experimental (model-independent) data on the short-lived fission products. Fissile isotopes of uranium, neptunium, plutonium and americium were irradiated under pulse conditions at the Washington State University 1 MW TRIGA reactor to achieve ~108 fissions. The samples were placed on a HPGe (high purity germanium) detector to begin counting in less than 3 minutes post irradiation. The samples were counted for various time intervals ranging from 5 minutes to 1 hour. The data was then analyzed to determine which radionuclides could be quantified and compared to the published fission yield data.

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

2009-11-01

91

Fission or Fossil: Life Cycle Assessment of Hydrogen Production  

Microsoft Academic Search

A comparative hybrid life cycle assessment (LCA) was conducted to evaluate two different methods for hydrogen production. The environmental impacts from nuclear assisted thermochemical water splitting are compared to hydrogen production from natural gas steam reforming with CO2 sequestration. The results show that the two methods have significantly different impacts. The nuclear alternative has lower impacts on global warming potential,

Christian Solli; Anders Hammer Stromman; Edgar G. Hertwich

2006-01-01

92

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

SciTech Connect

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

Malinauskas, A.P.

1985-01-01

93

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

94

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

95

Decontamination of actinides and fission products from stainless steel surfaces  

SciTech Connect

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

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

1996-04-01

96

ORNL studies of fission product release under LWR accident conditions  

SciTech Connect

High burnup Zircaloy-clad UO{sub 2} fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction.

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

1991-01-01

97

Assessment of fission product yields data needs in nuclear reactor applications  

SciTech Connect

Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

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

2012-07-01

98

Neutron Cross Section Covariances for Structural Materials and Fission Products  

SciTech Connect

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

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

2011-12-01

99

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

NASA Astrophysics Data System (ADS)

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

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

2004-05-01

100

Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS  

SciTech Connect

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport. (authors)

Kleinhietpass, Ingo D.; Unger, Hermann; Wagner, Hermann-Josef; Koch, Marco K. [Ruhr-University Bochum, Postfach 10 21 48, 44721 Bochum, (Germany)

2006-07-01

101

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

102

A systematics of fission product mass yields with 5 gaussian functions  

E-print Network

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

Katakura, J I

2003-01-01

103

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

104

Secretory production of ricinoleic acid in fission yeast Schizosaccharomyces pombe.  

PubMed

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

Yazawa, Hisashi; Kumagai, Hiromichi; Uemura, Hiroshi

2013-10-01

105

Core heatup and fission product release from an HTGR core in an LOFC accident. [AYERM code  

SciTech Connect

The AYERM code is a computer program which has been developed for the high-temperature gas-cooled reactor (HTGR) safety research program. It is a conjunction of the heat conduction code, AYER, and a set of special subroutines. This modified AYER code can predict the time-dependent release of volatile fission products from a reactor core during a hypothetical loss-of-forced-circulation (LOFC) accident. The computation scheme is based on the finite element method. The function of the AYER code is to compute the temperature distribution and the temperature history of a reactor during an LOFC accident. The subroutines perform two functions. One group of the subroutines provides the essential input data, such as the properties, configuration, initial and boundary conditions, etc., of the reactor core. The other group combines the computed instant local temperature with the fuel model parameters (i.e., the decay and release constants, and the irradiation history of the fuel) to perform the fission product release calculations.

Cort, G.E.; Fu, J.H.

1976-08-01

106

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

107

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

108

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

SciTech Connect

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

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

2013-03-01

109

Feasibility Study for an Active 238UF6 Gas Target for Photo-Fission Experiments  

NASA Astrophysics Data System (ADS)

A series of fission experiments in the actinide region has been started at the superconducting Darmstadt linear accelerator S-DALINAC. For detailed investigations on, e.g., the energy dependence of fission modes, the population of fission isomers, or even the search for parity non-conservation effects (PNC) in the photon-induced fission process of 238U, high luminosities are needed. Increasing target thickness reduces mass and angular resolutions. One possible solution is the utilization of an active gas target containing UF6. In order to test UF6 as an admixture to standard counting gases (e.g. argon) and to study its properties, an ionization chamber has been built at Technische Universität Darmstadt. After testing the chamber with pure argon as a counting gas to evaluate signal quality and to determine the drift velocity, gaseous UF6 was filled into the chamber in steps of one mass-percent uranium for each measurement, where both signal quality and drift velocity at different admixtures have been determined. Up to two percent of uranium in the counting gas one finds that the drift velocity increases with UF6 content, while overall a good signal quality and energy resolution of the ionization chamber is preserved.

Freudenberger, M.; Eckardt, C.; Göök, A.; Enders, J.; Hehner, J.; von Neumann-Cosel, P.; Oberstedt, A.; Oberstedt, S.; Simon, H.

110

Fission products from the damaged Fukushima reactor observed in Hungary.  

PubMed

Fission products, especially (131)I, (134)Cs and (137)Cs, from the damaged Fukushima Dai-ichi nuclear power plant (NPP) were detected in many places worldwide shortly after the accident caused by natural disaster. To observe the spatial and temporal variation of these isotopes in Hungary, aerosol samples were collected at five locations from late March to early May 2011: Institute of Nuclear Research, Hungarian Academy of Sciences (ATOMKI, Debrecen, East Hungary), Paks NPP (Paks, South-Central Hungary) as well as at the vicinity of Aggtelek (Northeast Hungary), Tapolca (West Hungary) and Bátaapáti (Southwest Hungary) settlements. In addition to the aerosol samples, dry/wet fallout samples were collected at ATOMKI, and airborne elemental iodine and organic iodide samples were collected at Paks NPP. The peak in the activity concentration of airborne (131)I was observed around 30 March (1-3 mBq m(-3) both in aerosol samples and gaseous iodine traps) with a slow decline afterwards. Aerosol samples of several hundred cubic metres of air showed (134)Cs and (137)Cs in detectable amounts along with (131)I. The decay-corrected inventory of (131)I fallout at ATOMKI was 2.1±0.1 Bq m(-2) at maximum in the observation period. Dose-rate contribution calculations show that the radiological impact of this event at Hungarian locations was of no considerable concern. PMID:24437973

Bihari, Árpád; Dezs?, Zoltán; Bujtás, Tibor; Manga, László; Lencsés, András; Dombóvári, Péter; Csige, István; Ranga, Tibor; Mogyorósi, Magdolna; Veres, Mihály

2014-01-01

111

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

112

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

113

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

SciTech Connect

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

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

2004-07-01

114

Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept  

SciTech Connect

The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.

White, J.R.; Burns, T.J.

1980-02-01

115

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

SciTech Connect

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

Johnson, I.; Johnson, C.E.

1987-01-01

116

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

NASA Astrophysics Data System (ADS)

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

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

2008-11-01

117

Towards a multiscale approach for assessing fission product behaviour in UN  

NASA Astrophysics Data System (ADS)

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

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

2013-11-01

118

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

119

Transient fission gas behavior in uranium nitride fuel under proposed space applications  

NASA Astrophysics Data System (ADS)

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 limited 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, Daniel L.

1991-12-01

120

Fission gas release and swelling in uranium–plutonium mixed nitride fuels  

Microsoft Academic Search

Two uranium–plutonium mixed nitride, (U,Pu)N, fuel pins with different He-gap width were irradiated at a linear heating rate 75 kW\\/m to 4.3% FIMA in the experimental fast reactor JOYO, and nondestructive and destructive post irradiation examinations were carried out. Fission gas release rates were about 3.3% and 5.2%, and swelling rates were about 1.8% and 1.6%\\/% FIMA. From the radial

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

2004-01-01

121

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

122

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

NASA Astrophysics Data System (ADS)

We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure 4He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the 4He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

Lewis, J. M.; Kelley, R. P.; Murer, D.; Jordan, K. A.

2014-07-01

123

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

SciTech Connect

We demonstrate the unambiguous detection of the fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium fusion neutron generator and a high pressure {sup 4}He gas fast neutron scintillation detector. The energy deposition by individual neutrons is quantified, and energy discrimination is used to differentiate the induced fission neutrons from the mono-energetic interrogation neutrons. The detector can discriminate between different incident neutron energies using pulse height discrimination of the slow scintillation component of the elastic scattering interaction between a neutron and the {sup 4}He atom. Energy histograms resulting from this data show the buildup of a detected fission neutron signal at higher energies. The detector is shown here to detect a unique fission neutron signal from a natural uranium sample during active interrogation with a (d, d) neutron generator. This signal path has a direct application to the detection of shielded nuclear material in cargo and air containers. It allows for continuous interrogation and detection while greatly minimizing the potential for false alarms.

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

2014-07-07

124

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

125

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

NASA Astrophysics Data System (ADS)

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

Dolan, David J.

126

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

SciTech Connect

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

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

2005-10-01

127

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

SciTech Connect

99Tc is a subsurface contaminant of interest at numerous federal, industrial, and international facilities. However, as a mono-isotopic fission product, 99Tc lacks the ability to be used as a signature to differentiate between the different waste disposal pathways that could have contributed to subsurface contamination at these facilities. Ruthenium fission-product isotopes are attractive analogues for the characterization of 99Tc sources because of their direct similarity to technetium with regard to subsurface mobility, and their large fission yields and low natural background concentrations. We developed an inductively coupled plasma mass spectrometry (ICP-MS) method capable of measuring ruthenium isotopes in groundwater samples and extracts of vadose zone sediments. Samples were analyzed directly on a Perkin Elmer ELAN DRC II ICP-MS after a single pass through a 1-ml bed volume of Dowex AG 50W-X8 100-200 mesh cation exchange resin. Precise ruthenium isotopic ratio measurements were achieved using a low-flow Meinhard-type nebulizer and long sample acquisition times (150,000 ms). Relative standard deviations of triplicate replicates were maintained at less than 0.5% when the total ruthenium solution concentration was 0.1 ng/ml or higher. Further work was performed to minimize the impact caused by mass interferences using the dynamic reaction cell (DRC) with O2 as the reaction gas. The aqueous concentrations of 96Mo and 96Zr were reduced by more than 99.7% in the reaction cell prior to injection of the sample into the mass analyzer quadrupole. The DRC was used in combination with stable-mass correction to quantitatively analyze samples containing up to 2-orders of magnitude more zirconium and molybdenum than ruthenium. The analytical approach documented herein provides an efficient and cost-effective way to precisely measure ruthenium isotopes and quantitate total ruthenium (natural vs. fission-product) in aqueous matrixes.

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

2006-05-09

128

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

SciTech Connect

This Fission Product Transport (FPT) Phenomena Identification and Ranking Technique (PIRT) report briefly reviews the high-temperature gas-cooled reactor (HTGR) FPT mechanisms and then documents the step-by-step PIRT process for FPT. The panel examined three FPT modes of operation: (1) Normal operation which, for the purposes of the FPT PIRT, established the fission product circuit loading and distribution for the accident phase. (2) Anticipated transients which were of less importance to the panel because a break in the pressure circuit boundary is generally necessary for the release of fission products. The transients can change the fission product distribution within the circuit, however, because temperature changes, flow perturbations, and mechanical vibrations or shocks can result in fission product movement. (3) Postulated accidents drew the majority of the panel's time because a breach in the pressure boundary is necessary to release fission products to the confinement. The accidents of interest involved a vessel or pipe break, a safety valve opening with or without sticking, or leak of some kind. Two generic scenarios were selected as postulated accidents: (1) the pressurized loss-of-forced circulation (P-LOFC) accident, and (2) the depressurized loss-of-forced circulation (D-LOFC) accidents. FPT is not an accident driver; it is the result of an accident, and the PIRT was broken down into a two-part task. First, normal operation was seen as the initial starting point for the analysis. Fission products will be released by the fuel and distributed throughout the reactor circuit in some fashion. Second, a primary circuit breach can then lead to their release. It is the magnitude of the release into and out of the confinement that is of interest. Depending on the design of a confinement or containment, the impact of a pressure boundary breach can be minimized if a modest, but not excessively large, fission product attenuation factor can be introduced into the release path. This exercise has identified a host of material properties, thermofluid states, and physics models that must be collected, defined, and understood to evaluate this attenuation factor. The assembled PIRT table underwent two iterations with extensive reorganization between meetings. Generally, convergence was obtained on most issues, but different approaches to the specific physics and transport paths shade the answers accordingly. The reader should be cautioned that merely selecting phenomena based on high importance and low knowledge may not capture the true uncertainty of the situation. This is because a transport path is composed of several serial linkages, each with its own uncertainty. The propagation of a chain of modest uncertainties can lead to a very large uncertainty at the end of a long path, resulting in a situation that is of little regulatory guidance.

Morris, Robert Noel [ORNL

2008-03-01

129

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

SciTech Connect

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

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

2008-04-14

130

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

131

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

132

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

133

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

SciTech Connect

Grain boundary (GB) separation as a mechanism for fission gas release (FGR), complementary to gas bubble interlinkage, has been experimentally observed in irradiated light water reactor fuel. However there has been limited effort to develop physics-based models incorporating this mechanism for the analysis of FGR. In this work, a computational study is carried out to investigate GB separation in UO2 fuel under the effect of gas bubble pressure and hydrostatic stress. A non-dimensional stress intensity factor formula is obtained through 2D axisymmetric analyses considering lenticular bubbles and Mode-I crack growth. The obtained functional form can be used in higher length-scale models to estimate the contribution of GB separation to FGR.

Pritam Chakraborty; Michael R. Tonks; Giovanni Pastore

2014-09-01

134

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

NASA Astrophysics Data System (ADS)

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

Yun, Di

135

Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel  

NASA Astrophysics Data System (ADS)

Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

1994-06-01

136

Nuclear Power from Fission Reactors. An Introduction.  

ERIC Educational Resources Information Center

The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

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

137

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

138

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 ...

O’Sullivan, Francis Martin

139

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.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

2013-11-01

140

Mass-yield distributions of fission products from 20, 32, and 45 MeV proton-induced fission of 232Th  

NASA Astrophysics Data System (ADS)

The yields of various fission products in the 19.6, 32.2, and 44.8 MeV proton-induced fission of 232Th have been determined by recoil catcher and an off-line ?-ray spectrometric technique using the BARC-TIFR Pelletron in India and MC-50 cyclotron in Korea. The mass-yield distributions were obtained from the fission product yield using the charge distribution corrections. The peak-to-valley (P/V) ratio of the present work and that of literature data for 232Th(p,f) and 238U(p,f) were obtained from the mass yield distribution. The present and the existing literature data for 232Th(p,f), 232Th(n,f), and 232Th( ?,f) at various energies were compared with those for 238U(p,f), 238U(n,f), and 238U( ?,f) to examine the probable nuclear structure effect. The role of Th-anomaly on the peak-to-valley ratio in proton-, neutron-, and photon-induced fission of 232Th was discussed with the similar data in 238U. On the other hand, the fine structure in the mass yield distributions of the fissioning systems at various excitation energies has been explained from the point of standard I and II asymmetric mode of fission besides the probable role of even-odd effect, A/ Z ratio, and fissility parameter.

Naik, H.; Goswami, A.; Kim, G. N.; Kim, K.; Suryanarayana, S. V.

2013-10-01

141

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

Microsoft Academic Search

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

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

2006-01-01

142

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

SciTech Connect

Fission gas bubble is one of evolving microstructures, which affect thermal mechanical properties such as thermo-conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phasefield model was developed to describe the evolution kinetics of intra-granular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nuclei size of the gas bubble and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter is predicted which is in a good agreement with experimental data.

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

2013-05-15

143

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

NASA Astrophysics Data System (ADS)

Fission gas bubbles are one of the evolving microstructures that affect thermal mechanical properties, such as thermal conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phase-field model was developed to describe the evolution kinetics of intragranular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nucleus size of gas bubbles and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter was predicted, which is in good agreement with experimental data.

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

2013-05-01

144

SOME ASPECTS OF THE BEHAVIOR OF FISSION PRODUCTS IN MOLTEN FLUORIDE REACTOR FUELS  

Microsoft Academic Search

0bservations are reported on the behavior of several fission product ; elements in molten NaF- ZrFâ-UFâ fuels, irradiated in capsule ; experiments, forcedconvection in-pile loop experiments, and in the Aircraft ; Reactor Experiment (ARE). The rare gases have been observed to escape readily ; from the fuels in dynamic tests, although in static tests the rate of escape is ;

M. T. Robinson; Brooksbank W. A. Jr. J Reynolds S. A; H. W. Wright; T. H. Handley

1958-01-01

145

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

Microsoft Academic Search

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

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

1983-01-01

146

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...

147

Ionization of the Atmosphere Due to Beta Particles Emitted by Fission Products  

Microsoft Academic Search

Persistent ionization of air at moderate heights, just below the E ; layer, after high-altitude nuclear detonations is predominantly due to radiations ; accompanying radioactive decay of fission products. The most important are BETA ; particles, which are restricted in their movements by the earth's magnetic field ; and thus create ionized clouds of high density in localized regions. A

S. Kownacki

1963-01-01

148

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

149

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

E-print Network

burnup, production history and the plutonium separation process used. Detailed understanding of the plutonium isotopic composition and fission product contaminant concentrations in separated plutonium would aid nuclear forensics activities aimed at source...

Osborn, Jeremy

2014-08-13

150

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

SciTech Connect

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

Blaise Collin

2014-09-01

151

Effects of High Target Atom Directed Velocity and Temperature on Interaction Rates for a Flowing Fissioning Gas  

Microsoft Academic Search

This dissertation represents an effort to develop a method to determine the effects of high target atom directed velocity and temperature on interaction rates for a flowing fissioning gas. Major emphasis is placed on the determination of the thermal neutron spectrum, microscopic neutron cross section behavior, and averaged macroscopic neutron cross section behavior. The system used as a model in

Claudio Luiz de Oliveira

1991-01-01

152

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

SciTech Connect

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

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

2014-04-07

153

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

154

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

SciTech Connect

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

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

1991-12-31

155

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

SciTech Connect

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

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

1991-01-01

156

Post irradiation examination of simulated fission product doped hyperstoichiometric mixed oxide fuel pins*1  

NASA Astrophysics Data System (ADS)

Two miniature fuel pins containing uranium-plutonium oxide with a hyperstoichiometric oxygen-to-metal ratio and selective fission product elements have been irradiated in the BR 2 reactor at Mol, Belgium, for two reactor cycles (46 days). One of the pins had a niobium metal coating on the inner cladding surface to act as oxygen getter. Both pins were subjected to a detailed examination by ceramography and electronprobe microanalysis. The results have been interpreted in the light of a recently published thermochemical model for the cladding attack. The very different oxygen potential environments in the two pins produced entirely different clad corrosion phenomena probably due to different cladding attack mechanisms. The niobium coating worked well in reducing the oxygen potential. However, there exists a draw back with niobium due to the formation of relatively stable intermetallic phases with noble metal fission products.

Götzmann, O.; Kleykamp, H.

1980-03-01

157

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

158

Fission-Product Separation Based on Room-Temperature Ionic Liquids  

SciTech Connect

The objectives of this project are (a) to synthesize new ionic liquids tailored for the extractive separation of Cs + and Sr 2+; (b) to select optimum macrocyclic extractants through studies of complexation of fission products with macrocyclic extractants and transport in new extraction systems based on ionic liquids; (c) to develop efficient processes to recycle ionic liquids and crown ethers; and (d) to investigate chemical stabilities of ionic liquids under strong acid, strong base, and high-level-radiation conditions.

Luo, Huimin

2006-11-15

159

Fission-Product Separation Based on Room-Temperature Ionic Liquids  

SciTech Connect

The objectives of this project are (a) to synthesize new ionic liquids tailored for the extractive separation of Cs + and Sr 2+; (b) to select optimum macrocyclic extractants through studies of complexation of fission products with macrocyclic extractants and transport in new extraction systems based on ionic liquids; (c) to develop efficient processes to recycle ionic liquids and crown ethers; and (d) to investigate chemical stabilities of ionic liquids under strong acid, strong base, and high-level-radiation conditions.

Luo, Huimin; Hussey, Charles L.

2005-09-30

160

Investigation on Correlations between Fractionation Effects and Particle Size Distributions in Fission Product Samples  

Microsoft Academic Search

After the third Chinese nuclear bomb test, radioactive aerosol was collected on fiber filters from the atmosphere at different altitudes. By comparing the measured p--decay curves with the theoretical F-decay curves the fractionation factor of the fresh fission products and the activity of 7U in the samples were determined. Furthermore, the samples were autoradiographed and the spot-size distribution and background

Wolfgang Peters; Dieter Paffrath

1970-01-01

161

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

162

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

SciTech Connect

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

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

1988-01-01

163

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

164

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

SciTech Connect

The objective of this project is to determine the diffusivity and chemical behavior of key fission products (ag, Cs, I. Te, Eu and Sr) through SiC and PyC both thermally, under irradiation, and under stress using FP introduction techniques that avoid the pitfalls of past experiments. The experimental approach is to create thin PyC0SiC couples containing the fission product to be studied embedded in the PyC layer. These samples will then be subjected to high temperature exposures in a vacuum and also to irradiation at high temperature, and last, to irradiation under stress at high temperature. The PyC serves as a host layer, providing a means of placing the fission product close to the SiC without damaging the SiC layer by its introduction or losing the FP during heating. Experimental measurements of grain boundary structure and distribution (EBSD, HRTEM, APT) will be used in the modeling efort to determine the qualitative dependence of FP diffusion coefficients on grain boundary orientation, temperature and stress.

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

2014-07-07

165

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

SciTech Connect

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

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

1988-01-01

166

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

Code of Federal Regulations, 2010 CFR

...2010-04-01 false (Item 1204) Oil and gas production, production prices and...Disclosure by Registrants Engaged in Oil and Gas Producing Activities § 229.1204 (Item 1204) Oil and gas production, production prices...

2010-04-01

167

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

Microsoft Academic Search

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

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

2010-01-01

168

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

Microsoft Academic Search

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

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

2010-01-01

169

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

Microsoft Academic Search

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

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

1979-01-01

170

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

NASA Astrophysics Data System (ADS)

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

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

2014-03-01

171

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

PubMed

An experiment was performed at the scientific neutron source FRM II in Garching to determine the cumulative antineutrino spectrum of the fission products of U238. Target foils of natural uranium were irradiated with a thermal and a fast neutron beam and the emitted ? spectra were recorded with a ?-suppressing electron telescope. The obtained ? spectrum of the fission products of U235 was normalized to the data of the magnetic spectrometer BILL. This method strongly reduces systematic errors in the U238 measurement. The ? spectrum of U238 was converted into the corresponding ?¯e spectrum. The final ?¯e spectrum is given in 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. PMID:24724646

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

2014-03-28

172

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

NASA Astrophysics Data System (ADS)

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

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

2013-05-01

173

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

SciTech Connect

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

James Stubbins

2012-12-19

174

Production Trends of Shale Gas Wells  

E-print Network

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

Khan, Waqar A.

2010-01-14

175

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

176

Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges  

SciTech Connect

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

Bibler, N

2005-04-05

177

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

178

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

179

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

180

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

SciTech Connect

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

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

1984-11-01

181

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

NASA Astrophysics Data System (ADS)

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

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

2013-07-01

182

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

NASA Astrophysics Data System (ADS)

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

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

2014-07-01

183

High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products  

SciTech Connect

An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ``engineered safety features,`` which, along with the use of high temperature capable materials further enhance its safety characteristics.

Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

1993-08-01

184

Immobilization of fission products in low-temperature ceramic waste forms  

SciTech Connect

Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics (CBPCs) for use in solidifying and stabilizing low-level mixed wastes. The focus of this work is development of CBPCs for use with fission-product wastes generated from high-level waste (HLW) tank cleaning or other decontamination and decommissioning activities. The volatile fission products such as Tc, Cs, and Sr removed from HLW need to be disposed of in a low-temperature immobilization system. Specifically, this paper reports on the solidification and stabilization of separated {sup 99}Tc from Los Alamos National Laboratory`s complexation-elution process. Using rhenium as a surrogate form technetium, we fabricated CBPC waste forms by acid-base reactions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with 35 wt.% waste loading. Standard leaching tests such as ANS 16.1 and PCT were conducted on the final waste forms. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water-immersion tests.

Singh, D.; Wagh, A.S.; Tlustochowicz, M.; Mandalika, V.

1997-01-01

185

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

SciTech Connect

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

Rest, J.

1995-08-01

186

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

SciTech Connect

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

Martin, R.C.

1990-01-01

187

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

188

Corrosion behavior of 9CrODS steel by simulated fission product cesium and tellurium  

NASA Astrophysics Data System (ADS)

Out-of-pile FCCI tests for 9CrODS steel were performed at 973 K by using simulated fission products Cs and Te under the oxygen potential in equilibrium with Fe/FeO and Cr/Cr2O3. Al2O3 powder were inserted to reduce a concentration of the Cs and Te in the system; its molar fraction is Cs:Te:Al2O3 = 1:1:1000. From EPMA and XRD analyses, Cr2O3 was formed at the most outer layer, which significantly suppressed the fission product corrosion. Cr2Te3 was also produced at the outer layer and interior of 9CrODS steel through liquid Te migration along grain boundaries. It was demonstrated the corrosion depth of 9CrODS steel is between PNC-FMS and PNC316, which were tested as reference. The Cs and Te assisted corrosion of 9CrODS steel was thermodynamically analyzed through the formation of Cs2O, Cs3CrO4, Cr2O3 and Cr2Te3.

Ukai, S.; Yamazaki, Y.; Oono, N.; Hayashi, S.

2013-09-01

189

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

SciTech Connect

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

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

1984-01-01

190

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

NASA Astrophysics Data System (ADS)

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

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

2013-06-01

191

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

SciTech Connect

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

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

2010-12-15

192

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

193

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

SciTech Connect

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

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

2014-05-01

194

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

SciTech Connect

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

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

2012-10-01

195

Seminar on Fission VI  

NASA Astrophysics Data System (ADS)

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

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

2008-04-01

196

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. Taín; F. Yermia; A. -A. Zakari-Issoufou

2012-09-13

197

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

PubMed

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

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

2012-11-16

198

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

199

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

200

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

SciTech Connect

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

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

2005-01-01

201

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

SciTech Connect

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

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

2005-03-15

202

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

203

Caspian Oil and Gas: Production and Prospects  

Microsoft Academic Search

Summary There is a likelihood of relatively large reserves of crude oil and natural gas in the Caspian Sea region, and a consequent large increase in oil and natural gas production from that area. Because diversity of energy sources and energy security are considerations in Congressional deliberations on energy policy, this prospect could play a role in such discussions. However,

Bernard A. Gelb

204

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

205

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

206

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

SciTech Connect

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

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

2007-04-27

207

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

SciTech Connect

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

Gilles Youinou; Andrea Alfonsi

2012-03-01

208

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

209

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

NASA Astrophysics Data System (ADS)

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

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

2014-12-01

210

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

SciTech Connect

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

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

2014-10-01

211

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

Microsoft Academic Search

The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels.

Gauntt; Randall O

2010-01-01

212

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

213

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

214

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

SciTech Connect

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

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

1983-09-01

215

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

216

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

217

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

218

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/?A·h and 21 MBq/?A·h for proton energies of 22 MeV and 40 MeV, respectively, energies that are available from many cyclotrons. With respect to 99Mo reactor based methods using uranium targets, the presented concept using 232Th does not pose proliferation concerns, transport of highly radioactive target materials can be reduced and unused cyclotron capacities could be exploited. Radiochemical target processing could be based on existing technologies of extraction of 99Mo from reactor irradiated 235U. The presented method could be used for co-production of other radioisotopes of medical interest such as 131I.

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

2012-05-01

219

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

220

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

221

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

222

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

223

Fission Xenon on Mars  

NASA Technical Reports Server (NTRS)

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

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

2002-01-01

224

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

225

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

226

Powering the World: Offshore Oil & Gas Production  

E-print Network

Powering the World: Offshore Oil & Gas Production Macondo post-blowout operations Tad Patzek;Talk Outline. . . Fuels that run the U.S. and world Drilling and fracturing primer Complexity and risks, not resources or energy potentially available Offshore fields will be producing an increasing portion of global

Patzek, Tadeusz W.

227

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

228

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

229

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

230

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

SciTech Connect

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

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

1993-10-01

231

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

Microsoft Academic Search

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

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

2009-01-01

232

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

Microsoft Academic Search

The electrolysis of burnup-simulated uranium nitride, UN, containing representative solid fission product elements (Mo, Pd, Nd) was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl from the view point of application of pyrochemical reprocessing to nitride fuel cycle. It was found from cyclic voltammetry and anodic polarization curve measurement that anodic dissolution of UN began at about

Takumi Satoh; Takashi Iwai; Yasuo Arai

2007-01-01

233

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

234

Delayed fission product gamma-ray transmission through low enriched uranium dioxide fuel pin lattices in air  

Microsoft Academic Search

The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray

Timothy H. Trumbull

2004-01-01

235

Automated analysis for large amount gaseous fission product gamma-scanning spectra from nuclear power plant and its data mining  

Microsoft Academic Search

Based on the Linssi database and UniSampo\\/Shaman software, an automated analysis platform has been setup for the analysis\\u000a of large amounts of gamma-spectra from the primary coolant monitoring systems of a CANDU reactor. Thus, a database inventory\\u000a of gaseous and volatile fission products in the primary coolant of a CANDU reactor has been established. This database is\\u000a comprised of 15,000

Weihua Zhang; Jarmo Ala-Heikkila; Kurt Ungar; Ian Hoffman; Ryan Lawrie

2010-01-01

236

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

237

Recent developments for an active UF6 gas target for photon-induced fission experiments  

NASA Astrophysics Data System (ADS)

Recent developments for an active uranium-hexafluoride-loaded gas target as well as results on the detector gas properties are presented. The gas of choice is a mixture of argon with small amounts of UF6. This contribution presents the experimental setup and focusses on the electron drift velocity with increasing UF6 content. A time-dependent decrease in electron drift velocity is observed in our setup.

Freudenberger, M.; Eckardt, C.; Enders, J.; Göök, A.; von Neumann-Cosel, P.; Oberstedt, A.; Oberstedt, S.

2013-12-01

238

Effects of High Target Atom Directed Velocity and Temperature on Interaction Rates for a Flowing Fissioning Gas  

NASA Astrophysics Data System (ADS)

This dissertation represents an effort to develop a method to determine the effects of high target atom directed velocity and temperature on interaction rates for a flowing fissioning gas. Major emphasis is placed on the determination of the thermal neutron spectrum, microscopic neutron cross section behavior, and averaged macroscopic neutron cross section behavior. The system used as a model in this study is a gas cavity region where a low density mixture of fissioning fuel and working fluid is in thermodynamic equilibrium at a high temperature and flowing at constant velocity. The gas cavity region is surrounded by a moderator which is the thermal neutron source. A brief review of gas core reactors concepts and neutron thermalization is presented. The development of these topics is traced historically, and the scope of the present work is established. The exact analytical development for the effective microscopic cross section and for the scattering kernel based on the works of Perkins and Oblow is presented. An approximate method to solve the problem for the conditions of completely entrained neutrons and uncollided neutrons is developed and is tested against the exact analytical solution for extreme conditions of temperature and directed velocity. The results from using the method are in agreement with the analytical solution. The method uses two different velocity frames and is suitable for implementation in a computer code. Results for the effective microscopic cross section calculated using this approach are tested against the only known and very limited results which are relevant to this problem and show good agreement. The developed approach is implemented in the computer code BRT-I, which solves the integral form of the neutron transport equation in the thermal energy region, following its adaptation to run in a personal computer. The modifications and adaptations included in BRT-I led to the creation of a new version of the code which is called BRT-PCM. The high directed velocity and temperature effects on the thermal neutron spectrum, on the averaged macroscopic cross sections, and on the average inverse neutron velocity are established and explained. From the results obtained, approximate analytical expressions to determine the thermal neutron spectrum and the microscopic and average macroscopic neutron cross sections within the range of studied temperatures and directed velocities are developed for the completely entrained and uncollided neutron conditions.

de Oliveira, Claudio Luiz

1991-02-01

239

Gas extrusion in natural products total synthesis.  

PubMed

The thermodynamic driving force from the release of a gaseous molecule drives a broad range of synthetic transformations. This review focuses on gas expulsion in key reactions within natural products total syntheses, selected from the past two decades. The highlighted examples survey transformations that generate sulfur dioxide, carbon dioxide, carbonyl sulfide, or nitrogen through polar, radical, pericyclic, photochemical, or organometallic mechanisms. Of particular interest are applications wherein the gas extrusion enables formation of a synthetically challenging motif, such as an unusually hindered or strained bond. PMID:22940671

Jiang, Xuefeng; Shi, Lei; Liu, Hui; Khan, Akbar H; Chen, Jason S

2012-11-14

240

Uranium dioxide films with xenon filled bubbles for fission gas behavior studies  

NASA Astrophysics Data System (ADS)

Electron beam evaporation and ion beam assisted deposition (IBAD) methods were utilized to fabricate depleted UO2 films and UO2 films with embedded Xe atoms, respectively. The films were fabricated at elevated temperature of 700 °C and also subsequently annealed at 1000 °C to induce grain growth and Xe atom redistribution. The goal of this work was to synthesize reference UO2 samples with controlled microstructures and Xe-filled bubble morphologies, without the effects attendant to rector irradiation-induced fission. Transmission electron microscopy (TEM) microstructural characterization revealed that fine Xe-filled bubbles nucleated in the as grown films and subsequent annealing resulted in noticeable bubble size increase. Reported results demonstrate the great potential IBAD techniques and UO2 films have for various areas of nuclear materials studies.

Usov, I. O.; Dickerson, R. M.; Dickerson, P. O.; Byler, D. D.; McClellan, K. J.

2014-09-01

241

EXTRACTION OF Am FROM NITRIC ACID BY CARBAMOYL-PHOSPHORYL EXTRACTANTS: THE INFLUENCE OF SUBSTITUENTS ON THE SELECTIVITY OF Am OVER Fe AND SELECTED FISSION PRODUCTS  

Microsoft Academic Search

A number of neutral extractants containing the P(0)(CH2)nC(0)N raolety were evaluated for their ability to extract Am from nitric acid and their selectivity for Am over Fe and selected fission products. Extractants containing alkoxy, alkyl, and aryl substltuents were evaluated. Tetrachloroethylene was used as a diluent. Fission products selected for study were Y, Zr, Mo, Tc, Ru, Rh, Pd, La,

E. Philip Horwltz; Kathleen A. Martin; Herbert Diamond; Louis Kaplan

1986-01-01

242

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

243

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

244

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

245

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

246

Investigations Into Devonian Shale Gas Production Mechanisms in Southern Ohio  

Microsoft Academic Search

Economic gas production from the Devonian Shale requires permeable pathways combined with matrix storage. These pathways may include fractures, bedding planes or silt layers. The Gas Research Institute is sponsoring a research project to evaluate the relationships these geologic features and productive gas flows have with the eventual aim of developing better exploration, stimulation and production strategies. This study will

T. W. Thompson; R. A. McBane; Gary Sitler; Jon Strawn; Mark Moody

1984-01-01

247

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

248

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

SciTech Connect

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

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

2014-11-01

249

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

E-print Network

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

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

2014-05-30

250

Spontaneous-fission decay properties and production cross-sections for the neutron-deficient nobelium isotopes formed in the 44, 48 Ca + 204, 206, 208 Pb reactions  

Microsoft Academic Search

:   Heavy-ion fusion reactions 48Ca + 204Pb and 44Ca + 208Pb leading to the same compound nucleus 252No* were run in attempts to produce new neutron-deficient spontaneous-fission isotopes of 249,250No using the electrostatic separator VASSILISSA. Production cross-sections for the spontaneous-fission activities with the\\u000a half-lives 5.6 and 54 ?s observed in these reactions are compared with the measured ones for the

A. V. Belozerov; M. L. Chelnokov; V. I. Chepigin; T. P. Drobina; V. A. Gorshkov; A. P. Kabachenko; O. N. Malyshev; I. M. Merkin; Yu. Ts. Oganessian; A. G. Popeko; R. N. Sagaidak; A. I. Svirikhin; A. V. Yeremin; G. Berek; I. Brida; Š. Šáro

2003-01-01

251

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

SciTech Connect

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

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

2014-03-01

252

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

NASA Astrophysics Data System (ADS)

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

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

2014-08-01

253

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

SciTech Connect

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

Suwardi; Dewayatna, W.; Briyatmoko, B. [Center for Nuclear Fuel Technology - National Nuclear Energy Agency, Puspiptek Tangerang - 15310 (Indonesia)

2012-06-06

254

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

NASA Astrophysics Data System (ADS)

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

Suwardi, Dewayatna, W.; Briyatmoko, B.

2012-06-01

255

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

Microsoft Academic Search

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

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

1978-01-01

256

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

257

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

SciTech Connect

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

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

2012-12-01

258

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

SciTech Connect

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

Slater, C.O.

1990-05-01

259

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

SciTech Connect

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

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

1987-01-01

260

Gas lift design and production optimisation offshore Trinidad  

SciTech Connect

By means of a variety of field examples, this paper describes how increased production rates were obtained from gas lift wells. These results were achieved through a wide range of activities including, special training for production operators, optimising gas injection rates, modifying surface piping systems, identifying and replacing defective wireline-retrievable gas lift valves, and improving gas lift design techniques. A major modification of a standard gas lift design technique is discussed in detail. The modification optimizes the depth of gas injection throughout the life of a well. An empirically derived chart, which relates valve spacing to the productivity index of a well, is also presented.

Laing, C.M.

1986-01-01

261

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. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, "ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology," Nuclear Data Sheets 107, 2931 (2006)].

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

2011-12-01

262

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

263

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

SciTech Connect

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

Hikaru Hiruta; Gilles Youinou

2013-09-01

264

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

SciTech Connect

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

Myers, B.F.

1995-09-01

265

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

266

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

NASA Technical Reports Server (NTRS)

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

Lustig, P. H.

1973-01-01

267

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

PubMed

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

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

2014-01-01

268

30 CFR 206.174 - How do I value gas production when an index-based method cannot be used?  

Code of Federal Regulations, 2010 CFR

...any other gas production that cannot be...as well as gas plant products, and...Minimum value of production. (1) For...gas, and gas plant products valued...minimum value of production for each gas plant product is...

2010-07-01

269

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

270

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

271

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

PubMed

Ionizing radiation is known to increase intracellular level of reactive oxygen species (ROS) through mitochondrial dysfunction. Although it has been as a basis of radiation-induced genetic instability, the mechanism involving mitochondrial dysfunction remains unclear. Here we studied the dynamics of mitochondrial structure in normal human fibroblast like cells exposed to ionizing radiation. Delayed mitochondrial O(2)(-) production was peaked 3 days after irradiation, which was coupled with accelerated mitochondrial fission. We found that radiation exposure accumulated dynamin-related protein 1 (Drp1) to mitochondria. Knocking down of Drp1 expression prevented radiation induced acceleration of mitochondrial fission. Furthermore, knockdown of Drp1 significantly suppressed delayed production of mitochondrial O(2)(-). Since the loss of mitochondrial membrane potential, which was induced by radiation was prevented in cells knocking down of Drp1 expression, indicating that the excessive mitochondrial fission was involved in delayed mitochondrial dysfunction after irradiation. PMID:22005465

Kobashigawa, Shinko; Suzuki, Keiji; Yamashita, Shunichi

2011-11-01

272

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

SciTech Connect

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

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

2011-01-01

273

Thermal reactor. [liquid silicon production from silane gas  

NASA Technical Reports Server (NTRS)

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

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

1982-01-01

274

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

E-print Network

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

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

2015-01-01

275

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

E-print Network

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

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

2015-02-10

276

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

277

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

NASA Astrophysics Data System (ADS)

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

Bull, Geoffrey R.

278

Cryogenic production of ammonia synthesis gas  

Microsoft Academic Search

An improved cryogenic separation process is provided for forming a substantially CO-free and lower hydrocarbon-free hydrogen-rich gas, suitable for use in producing an ammonia synthesis gas, from a hydrogen-containing gas stream also containing carbon monoxide and lower hydrocarbon impurities, wherein the hydrogen containing gas is subject to a two-stage autorefrigerated cryogenic flash treatment to remove concentrated methane-containing and carbon monoxide-containing

Traficante

1985-01-01

279

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

280

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

281

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

282

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

PubMed

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

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

2011-01-01

283

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

PubMed

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

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

2014-07-01

284

Isomers in Fission Fragments  

SciTech Connect

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

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

2009-01-28

285

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

SciTech Connect

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

Henager, Charles H.; Jiang, Weilin

2014-11-01

286

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

287

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

288

Gas production and migration in landfills and geological materials  

NASA Astrophysics Data System (ADS)

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

Nastev, Miroslav; Therrien, René; Lefebvre, René; Gélinas, Pierre

2001-11-01

289

Offshore LNG (liquefied natural gas) production and storage systems  

Microsoft Academic Search

A barge, outfitted with gas liquefaction processing equipment and liquefied natural gas (LNG) storage tanks, is suggested as a possible way to exploit remote offshore gas production. A similar study with a barge-mounted methanol plant was conducted several years ago, also using remote offshore feed gas. This barge-mounted, LNG system is bow-moored to a single point mooring through which feed

Barden

1982-01-01

290

An off-line method to characterize the fission product release from uranium carbide-target prototypes developed for SPIRAL2 project  

NASA Astrophysics Data System (ADS)

In the context of radioactive ion beams, fission targets, often based on uranium compounds, have been used for more than 50 years at isotope separator on line facilities. The development of several projects of second generation facilities aiming at intensities two or three orders of magnitude higher than today puts an emphasis on the properties of the uranium fission targets. A study, driven by Institut de Physique Nucléaire d'Orsay (IPNO), has been started within the SPIRAL2 project to try and fully understand the behavior of these targets. In this paper, we have focused on five uranium carbide based targets. We present an off-line method to characterize their fission product release and the results are examined in conjunction with physical characteristics of each material such as the microstructure, the porosity and the chemical composition.

Hy, B.; Barré-Boscher, N.; Özgümüs, A.; Roussière, B.; Tusseau-Nenez, S.; Lau, C.; Cheikh Mhamed, M.; Raynaud, M.; Said, A.; Kolos, K.; Cottereau, E.; Essabaa, S.; Tougait, O.; Pasturel, M.

2012-10-01

291

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

292

Fission gas behavior in mixed-oxide fuel during transient overpower and simulated loss-of-flow tests  

Microsoft Academic Search

A portion of the fission gases (Xe, Kr) generated during steady state irradiation of an FBR mixed-oxide fuel is retained within its microstructure, especially in the cooler regions of the fuel. Information on the behavior of these retained gases relative to their microstructural relocation and release during off-normal power and coolant flow conditions is important to the analysis and analytic

E. H. Randklev; C. A. Hinman

1979-01-01

293

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

Microsoft Academic Search

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

G Chen; J Andries; Z Luo; H Spliethoff

2003-01-01

294

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

SciTech Connect

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

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

1994-12-31

295

Measuring micro-organism gas production  

NASA Technical Reports Server (NTRS)

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

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

1973-01-01

296

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

DOEpatents

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

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

1995-01-01

297

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

298

Isotopic production cross sections and recoil velocities of spallation-fission fragments in the reaction U238(1AGeV)+d  

NASA Astrophysics Data System (ADS)

Fission fragments of 1AGeVU238 nuclei interacting with a deuterium target have been investigated with the Fragment Separator (FRS) at Gesellschaft für Schwerionenforschung (GSI) by measuring their isotopic production cross sections and velocities. Results, along with those obtained recently for spallation-evaporation fragments, provide a comprehensive analysis of the spallation nuclear productions in this reaction. Details about the experimental performance, data reduction and results are presented.

Pereira, J.; Benlliure, J.; Casarejos, E.; Armbruster, P.; Bernas, M.; Boudard, A.; Czajkowski, S.; Enqvist, T.; Legrain, R.; Leray, S.; Mustapha, B.; Pravikoff, M.; Rejmund, F.; Schmidt, K.-H.; Stéphan, C.; Taïeb, J.; Tassan-Got, L.; Volant, C.; Wlazlo, W.

2007-01-01

299

Improving oil and gas production with the Beam-Mounted Gas Compressor  

SciTech Connect

This paper explores parameters involved in and advantages obtained by use of the Beam-Mounted Gas Compressor (BMGC), a single-acting gas compressor operated by the walking beam of a rod pumping unit. Its main function is to draw gas from the casing side of an oil well and to discharge the gas into the flow line. By doing so, the BMGC operation actually reduces the backpressure on the formation face, thus allowing additional oil to enter the wellbore for production.

Al-Khatib, A.M.

1984-02-01

300

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

SciTech Connect

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

McGuire, P.L.

1981-01-01

301

Calculated spectra of antineutrinos from the fission products of 235U, 238U, and 239Pu, and antineutrino-induced reactions  

NASA Astrophysics Data System (ADS)

The theoretical spectra of antineutrinos from the beta decays of the products of fissioning 235U, 238U, and 239Pu were recalculated using recent compilations of the level structure, beta branching ratios, and fission yields. In addition, a very recent semiempirical mass formula, particularly designed to achieve local accuracy on the (N, Z) plane, was used to calculate the beta Q values of isotopes of unknown decay schemes. Recent decay systematics, far from the beta stability line, have encouraged us to assume that most of the levels of the isotopes of a given Z, differing by ?N=2, are very similar in order and in energy. The beta spectrum from equilibrium, thermal fission products of 235U was calculated and compared with existing experimental data. Theoretical predictions are presented for the inverse beta decay of the proton, the neutral current disintegration of the deuteron, and the excitation of 7Li by antineutrinos. Existing data from the elastic scattering of electrons by antineutrinos are reanalyzed with the present antineutrino spectra. The result is that these data agree with the predictions of the Weinberg-Salam model with (0.24<2?W<0.31) and do not agree with the predictions of the Feynman-Gell-Mann theory. RADIOACTIVITY, FISSION 235U, 238U, 239Pu; antineutrino and beta spectra calculated in secular equilibrium. ?¯ for ?¯e(p, n)?+, ?¯e(d, pn)?¯e, ?¯e(7Li, 7Li*)?¯e, and ?¯e(e-, e-)?¯e.

Avignone, F. T., III; Greenwood, Z. D.

1980-08-01

302

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

303

Ground movements associated with gas hydrate production. Final report  

Microsoft Academic Search

This report deals with a study directed towards a modeling effort on production related ground movements and subsidence resulting from hydrate dissociation. The goal of this research study was to evaluate whether there could be subsidence related problems that could be an impediment to hydrate production. During the production of gas from a hydrate reservoir, it is expected that porous

H. J. Siriwardane; B. Kutuk

1992-01-01

304

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

SciTech Connect

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

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

2011-01-01

305

Strategies for gas production from oceanic Class 3 hydrateaccumulations  

SciTech Connect

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

Moridis, George J.; Reagan, Matthew T.

2007-05-01

306

New economics of natural gas production in the Appalachian states  

Microsoft Academic Search

Since at least 1920, production of natural gas in the Appalachian states has fluctuated between 400 and 500 billion cu ft\\/yr. Among the factors limiting expansion in drilling, recompletion of old wells, and introduction of new technology is the adverse wellhead prices paid to producers by purchasers under regulations applied by the federal government under the US Natural Gas Act

Jaworek

1979-01-01

307

Forecasting long-term gas production Luis Cueto-Felguerosoa  

E-print Network

and hydraulic fracturing. Hori- zontal drilling enhances the spatial access to the hydrocarbon resource formations. Current shale gas production relies on two quickly evolving technologies: horizon- tal drilling by increasing the length of a single well within the gas-bearing shale. Hydraulic fracturing, or "fracking" (9

Patzek, Tadeusz W.

308

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

309

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

SciTech Connect

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

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

2012-04-11

310

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

311

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

312

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

313

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

314

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

SciTech Connect

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

Satoh, Takumi; Iwai, Takashi; Arai, Yasuo [Japan Atomic Energy Agency, Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki-ken, 311-1393 (Japan)

2007-07-01

315

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

316

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

USGS Publications Warehouse

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

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

2012-01-01

317

Using tea as an artificial urine in a Canadian performance testing program for fission/activation products.  

PubMed

In recent years, the National Calibration Reference Centre for Bioassay and In Vivo Monitoring (NCRC) at the Radiation Protection Bureau (RPB), Health Canada, has been conducting investigations with black tea to develop a matrix that can be used to replace urine in each of the following performance testing programs (PTP): (1) tritium, (2) carbon-14, (3) the DUAL (i.e., 3H/14C), and (4) fission/activation products (F/AP). A 1% tea solution with thimerosal, which had worked successfully for tritium, carbon-14, and the DUAL, was selected and tested for the F/AP PTP because of its similarity to urine in color and UV-VIS spectra. However, application of this tea to samples of the F/AP program containing 133Ba, 137Cs, 57Co, and 60Co produced precipitates, which was an unexpected result. Further experiments showed that replacement of thimerosal with an alcohol at about 5% eliminated the precipitation problem. The alcohol can be ethanol, methanol, or isopropanol. In the experiments, the 1% tea, preserved with alcohol, remained clear and stable for at least 100 d. The duration of each PTP for the NCRC is limited to 90 d. Application of the CNSC S-106 regulatory standard to the tea produced acceptable accuracy and precision results. It was concluded that a suitable tea matrix for the F/AP program had been found. PMID:25353238

Daka, Joseph N; Moodie, Gerry; DiNardo, Anthony; Kramer, Gary H

2014-12-01

318

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

SciTech Connect

A ThermoFisher 'Triton' multi-collector thermal ionization mass spectrometer (MC-TIMS) was evaluated for trace and ultra-trace level isotoperatioanalysis 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 (10{sup 4} atoms to 10{sup 5} atoms) for {sup 239-242+244}Pu, {sup 233+236}U, {sup 241-243}Am, {sup 89,90}Sr, and {sup 134,135,137}Cs, and {le} 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 x 10{sup 6} or better using a SEM are reported here. Precisions of RSD {approx} 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.

Bürger, Stefan [New Brunswick Laboratory, Argonne, IL; Riciputi, Lee R [Los Alamos National Laboratory (LANL); Bostick, Debra A [ORNL; Turgeon, Steven [University of Alberta, Edmondton, Canada; McBay, Eddie H [ORNL; Lavelle, Mark [ORNL

2009-01-01

319

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.

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

2009-09-01

320

Development of a multi-layer diffusion couple to study fission product transport in ?-SiC  

NASA Astrophysics Data System (ADS)

A multi-layer diffusion couple was designed to study fission product diffusion behavior while avoiding the pitfalls of direct ion implantation. Thin films of highly anisotropic pyrolytic carbon (PyC) were deposited onto CVD ?-SiC substrates. The PyC films were subsequently implanted with 400 keV silver, cesium, strontium, europium, or iodine at 22 °C to a dose of 1016 cm-2, such that the implanted species resided wholly within the PyC layer. The samples were then coated with 50 nm of SiC via plasma enhanced CVD (PECVD) to retain the implanted species during post-deposition annealing treatments. The design allows for high spatial resolution tracking of the implanted specie using Rutherford backscattering spectrometry. Annealing at 1100 °C for 10 h resulted in retention of 100% of implanted cesium, strontium, europium and iodine, and 70% of silver. This diffusion couple design provides the opportunity to determine diffusion coefficients of FPs in PyC and SiC and the role of the PyC/SiC interface in FP transport.

Dwaraknath, S.; Was, G. S.

2014-01-01

321

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

322

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

323

Application of Computer-Based Operator Instruction System to Plant Diagnosis on Fission Product Transport and Release in Nuclear Power Plants  

Microsoft Academic Search

A computer-based operator instruction system (COINS) for diagnosing fission product (FP) transport and release in nuclear power plants (NPPs) is applied to plant diagnosis in combination with the computational code “SACHET”, which evaluates the dynamic FP inventories in the multiple compartment system of pressurized water reactor (PWR) plants.The COINS can be described in the most general way as a computer-based

Hideki KODAIRA; Shunsuke KONDO; Yasumasa TOGO

1985-01-01

324

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

325

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

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

326

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

SciTech Connect

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

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

2010-02-26

327

US production of natural gas from tight reservoirs  

SciTech Connect

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

Not Available

1993-10-18

328

Report on possible routes to breakdown products of mustard gas  

SciTech Connect

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

Luman, F.M.

1983-10-18

329

Chemical and Physical Properties of Dry Flue Gas Desulfurization Products  

Microsoft Academic Search

be out of compliance without remedial action. This prob- lemhasspurredthedevelopmentofvarioustypesofscrub- Beneficial and environmentally safe recycling of flue gas desulfur- bing processes to convert SO2 from flue gases into solid ization (FGD) products requires detailed knowledge of their chemical and physical properties. We analyzed 59 dry FGD samples collected products for disposal or beneficial reuse. These FGD from 13 locations representing

David A. Kost; Jerry M. Bigham; Richard C. Stehouwer; Joel H. Beeghly; Randy Fowler; Samuel J. Traina; William E. Wolfe; Warren A. Dick

2005-01-01

330

Investigations into devonian shale gas production mechanisms in southern Ohio  

SciTech Connect

Economic gas production from the Devonian Shale requires permeable pathways combined with matrix storage. These pathways may include fractures, bedding planes or silt layers. The Gas Research Institute is sponsoring a research project to evaluate the relationships these geologic features and productive gas flows have with the eventual aim of developing better exploration, stimulation and production strategies. This study will investigate wells throughout the basin, with a wide range of potential productivity in order to determine what variations in geology result in marginal versus good producers. As a first phase of this project, detailed data on the geology and reservoir parameters are being collected on four wells in southern Ohio, which are being recompleted in the Devonian Shale. Two additional wells have been investigated in a pilot study, the remainder are currently under investigation.

Thompson, T.W.; McBane, R.A.; Moody, M.; Sitler, G.; Strawn, J.A.

1984-09-01

331

Water Resources and Natural Gas Production from the Marcellus Shale  

USGS Publications Warehouse

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

Soeder, Daniel J.; Kappel, William M.

2009-01-01

332

On-Board Hydrogen Gas Production System For Stirling Engines  

DOEpatents

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

Johansson, Lennart N. (Ann Arbor, MI)

2004-06-29

333

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

334

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

335

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

336

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

337

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

E-print Network

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

338

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

339

Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities  

NASA Astrophysics Data System (ADS)

The radionuclide 125Sb is a long-lived fission product, which decays to 125Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

Rajput, M. U.; Ali, N.; Hussain, S.; Mujahid, S. A.; MacMahon, D.

2012-04-01

340

Greenhouse gas budgets of crop production current  

E-print Network

of agricultural output 13 2.3.3 Soil N2 O emissions from crop residue incorporation and N-fixing crops 13 2.4 CH4 4.7.3 Residue management 41 4.8 Mitigation potential of agronomy measures 42 4.8.1 Catch crops and increased cover 42 4.8.2 Crop selection and rotation 42 4.9 Mitigation potential in rice production 43 4

Levi, Ran

341

30 CFR 250.1629 - Additional production and fuel gas system requirements.  

Code of Federal Regulations, 2010 CFR

...false Additional production and fuel gas system requirements. 250.1629...DEPARTMENT OF THE INTERIOR OFFSHORE OIL AND GAS AND SULPHUR OPERATIONS IN THE OUTER...1629 Additional production and fuel gas system requirements. (a)...

2010-07-01

342

Natural gas production from hydrate dissociation: An axisymmetric model  

SciTech Connect

This paper describes an axisymmetric model for natural gas production from the dissociation of methane hydrate in a confined reservoir by a depressurizing well. During the hydrate dissociation, heat and mass transfer in the reservoir are analyzed. The system of governing equations is solved by a finite difference scheme. For different well pressures and reservoir temperatures, distributions of temperature and pressure in the reservoir, as well as the natural gas production from the well are evaluated. The numerical results are compared with those obtained by a linearization method. It is shown that the gas production rate is a sensitive function of well pressure. The simulation results are compared with the linearization approach and the shortcomings of the earlier approach are discussed.

Ahmadi, G. (Clarkson Univ., Pottsdam, NY); Ji, Chuang (Clarkson Univ., Pottsdam, NY); Smith, D.H.

2007-08-01

343

Engineering analysis of biomass gasifier product gas cleaning technology  

SciTech Connect

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

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

1986-08-01

344

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.

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

1986-01-01

345

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

346

Bimodal fission  

SciTech Connect

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

Hulet, E.K.

1989-04-19

347

Gas production from oceanic Class 2 hydrate accumulations  

SciTech Connect

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

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

2007-02-01

348

Some modern notions on oil and gas reservoir production regulation  

SciTech Connect

The historic rhetoric of oil and gas reservoir production regulations has been burdened with misconceptions. One was that most reservoirs are rate insensitive. Another was that a reservoir's decline is primarily a function of reservoir mechaism rather than a choice unconstrained by the laws of physics. Relieved of old notions like these, we introduce some modern notions, the most basic being that production regulation should have the purpose of obtaining the highest value from production per irreversible diminution of thermodynamically available energy. The laws of thermodynamics determine the available energy. What then is value. Value may include contributions other than production per se and purely monetary economic outcomes.

Lohrenz, J.; Monash, E.A.

1980-05-21

349

Hazardous Gas Production by Alpha Particles  

SciTech Connect

This project focused on the production of hazardous gases in the radiolysis of solid organic matrices, such as polymers and resins, that may be associated with transuranic waste material. Self-radiolysis of radioactive waste is a serious environmental problem because it can lead to a change in the composition of the materials in storage containers and possibly jeopardize their integrity. Experimental determination of gaseous yields is of immediate practical importance in the engineering and maintenance of containers for waste materials. Fundamental knowledge on the radiation chemical processes occurring in these systems allows one to predict outcomes in materials or mixtures not specifically examined, which is a great aid in the management of the variety of waste materials currently overseen by Environmental Management.

Jay A. LaVerne, Principal Investigator

2001-11-26

350

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

351

Producing Ammonium Sulfate from Flue Gas Desulfurization By-Products  

Microsoft Academic Search

Emission control technologies using flue gas desulfurization (FGD) have been widely adopted by utilities burning high-sulfur fuels. However, these technologies require additional equipment, greater operating expenses, and increased costs for landfill disposal of the solid by-products produced. The financial burdens would be reduced if successful high-volume commercial applications of the FGD solid by-products were developed. In this study, the technical

Mei-In M. Chou; Joshua A. Bruinius; Vincent Benig; Sheng-Fu J. Chou; Ronald H. Carty

2005-01-01

352

Gas Mixtures and Ozone Production in an Electrical Discharge  

Microsoft Academic Search

The quantitative production of ozone (O3) with N2, O2, and Ar gas mixtures in an atmospheric pressure corona discharge (CD) is investigated. A five-part model is presented that explores the discharge conditions needed for optimum ozone production. One part of the model is the well-known relationship that correlates the discharge's voltage, frequency, gap, dielectric material, etc with the generator's yield.

Thomas J. Manning; Jerry Hedden

2001-01-01

353

Pumps, refracturing hike production from tight shale gas wells  

Microsoft Academic Search

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

S. R. Reeves; W. K. Morrisson; D. G. Hill

1993-01-01

354

Entropy Production and Thermal Conductivity of A Dilute Gas  

E-print Network

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

Yong-Jun Zhang

2011-02-16

355

BUILDING MATERIALS MADE FROM FLUE GAS DESULFURIZATION BY-PRODUCTS  

SciTech Connect

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

Michael W. Grutzeck; Maria DiCola; Paul Brenner

2006-03-30

356

Depressurization-induced gas production from Class 1 hydratedeposits  

SciTech Connect

Class 1 hydrate deposits are characterized by ahydratebearing layer underlain by a two-phase zone involving mobile gas.Two kinds of deposits are investigated. The first involves water andhydrate in the hydrate zone (Class 1W), while the second involves gas andhydrate (Class 1G). We introduce new models to describe the effect of thepresence of hydrates on the wettability properties of porous media. Wedetermine that large volumes of gas can be readily produced at high ratesfor long times from Class 1 gas hydrate accumulations by means ofdepressurization-induced dissociation using conventional technology.Dissociation in Class 1W deposits proceeds in distinct stages, while itis continuous in Class 1G deposits. To avoid blockage caused by hydrateformation in the vicinity of the well, wellbore heating is a necessity inproduction from Class 1 hydrates. Class 1W hydrates are shown tocontribute up to 65 percent of the production rate and up to 45 percentof the cumulative volume of produced gas; the corresponding numbers forClass 1G hydrates are 75 percent and 54 percent. Production from bothClass 1W and Class 1G deposits leads to the emergence of a seconddissociation front (in addition to the original ascending hydrateinterface) that forms at the top of the hydrate interval and advancesdownward. Inboth kinds of deposits, capillary pressure effects lead tohydrate lensing, i.e., the emergence of distinct banded structures ofalternating high-low hydrate saturation, which form channels and shellsand have a significant effect on production.

Moridis, George J.; Kowalsky, Michael B.; Pruess, Karsten

2005-11-01

357

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

358

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

359

Environmental Compliance for Oil and Gas Exploration and Production  

SciTech Connect

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

Hansen, Christine

1999-10-26

360

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

361

PRODUCTION OF SYNTHETIC NATURAL GAS FROM BIOMASS - PROCESS INTEGRATED DRYING  

Microsoft Academic Search

Opportunities for process integrated feedstock drying in connection with the production of synthetic natural gas (SNG) from wet biomass via indirect gasification are investigated in this study. Drying is a very energy-intensive process step - corresponding to about 10% of the dry fuel lower heating value for woody biomass. Process integrated drying offers opportunities for reducing the external energy supply

Stefan Heyne; Simon Harvey

362

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

363

Microscopic description of complex nuclear decay: Multimodal fission  

NASA Astrophysics Data System (ADS)

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.

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

2009-07-01

364

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.

Staszczak, A; Dobaczewski, J; Nazarewicz, W

2009-01-01

365

Microscopic description of complex nuclear decay: Multimodal fission  

SciTech Connect

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.

Staszczak, A.; Baran, A. [Institute of Physics, Maria Curie-Sklodowska University, pl. M. Curie-Sklodowskiej 1, PL-20-031 Lublin (Poland); Department of Physics and Astronomy, University of Tennessee Knoxville, Tennessee 37996 (United States); Physics Division, Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, Tennessee 37831 (United States); Dobaczewski, J. [Institute of Theoretical Physics, University of Warsaw, ul. Hoza 69, PL-00-681 Warsaw (Poland); Department of Physics, P. O. Box 35 (YFL), FI-40014 University of Jyvaeskylae (Finland); Nazarewicz, W. [Department of Physics and Astronomy, University of Tennessee Knoxville, Tennessee 37996 (United States); Physics Division, Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, Tennessee 37831 (United States); Institute of Theoretical Physics, University of Warsaw, ul. Hoza 69, PL-00-681 Warsaw (Poland)

2009-07-15

366

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

367

Fission meter  

DOEpatents

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

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

2012-04-10

368

Analytical Modeling of Shale Hydraulic Fracturing and Gas Production  

NASA Astrophysics Data System (ADS)

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

Xu, W.

2012-12-01

369

Relative water and gas permeability for gas production from hydrate-bearing sediments  

NASA Astrophysics Data System (ADS)

water and gas permeability equations are important for estimating gas and water production from hydrate-bearing sediments. However, experimental or numerical study to determine fitting parameters of those equations is not available in the literature. In this study, a pore-network model is developed to simulate gas expansion and calculate relative water and gas permeability. Based on the simulation results, fitting parameters for modified Stone equation are suggested for a distributed hydrate system where initial hydrate saturations range from Sh = 0.1 to 0.6. The suggested fitting parameter for relative water permeability is nw ? 2.4 regardless of initial hydrate saturation while the suggested fitting parameter for relative gas permeability is increased from ng = 1.8 for Sh = 0.1 to ng = 3.5 for Sh = 0.6. Results are relevant to other systems that experience gas exsolution such as pockmark formation due to sea level change, CO2 gas formation during geological CO2 sequestration, and gas bubble accumulation near the downstream of dams.

Mahabadi, Nariman; Jang, Jaewon

2014-06-01

370

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

371

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

...regarding oil and natural gas production by addressing issues such...equipment lifecycle analysis, production safety systems, subsurface safety devices...for operating dry tree and subsea tree production systems on the Outer...

2013-09-27

372

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

373

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

374

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

375

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

376

Effects of gas chamber geometry and gas flow on the neutron production in a fast plasma focus neutron source  

NASA Astrophysics Data System (ADS)

This work reports that gas chamber geometry and gas flow management substantially affect the neutron production of a repetitive fast plasma focus. The gas flow rate is the most sensitive parameter. An appropriate design of the gas chamber combined with a suitable flow-rate management can lead to improvements in the neutron production of one order of magnitude working in a fast repetitive mode.

Tarifeño-Saldivia, Ariel; Soto, Leopoldo

2014-12-01

377

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

378

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

USGS Publications Warehouse

The current paper complements the Moridis et al. (2009) review of the status of the effort toward commercial gas production from hydrates. We aim to describe the concept of the gas-hydrate (GH) petroleum system; to discuss advances, requirements, and suggested practices in GH prospecting and GH deposit characterization; and to review the associated technical, economic, and environmental challenges and uncertainties, which include the following: accurate assessment of producible fractions of the GH resource; development of methods for identifying suitable production targets; sampling of hydrate-bearing sediments (HBS) and sample analysis; analysis and interpretation of geophysical surveys of GH reservoirs; well-testing methods; interpretation of well-testing results; geomechanical and reservoir/well stability concerns; well design, operation, and installation; field operations and extending production beyond sand-dominated GH reservoirs; monitoring production and geomechanical stability; laboratory investigations; fundamental knowledge of hydrate behavior; the economics of commercial gas production from hydrates; and associated environmental concerns. ?? 2011 Society of Petroleum Engineers.

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

2011-01-01

379

Sodic soils reclaimed with by-product from flue gas desulfurization: corn production and soil quality  

Microsoft Academic Search

Interest is growing in the use of by-product from flue gas desulfurization (FGD) to reclaim sodic soils by controlling the pH and excessive Na+. This study evaluated the effects on corn (Zea mays) production and pH and electrical conductivity (EC) of calcareous sodic soil during four times of cultivation when the by-product was applied once at the first cultivation (Study

S Chun; M Nishiyama; S Matsumoto

2001-01-01

380

Flue gas desulfurization by-products additions to acid soil: alfalfa productivity and environmental quality  

Microsoft Academic Search

Flue gas desulfurization (FGD) by-products are created when coal is burned and SO2 is removed from the flue gases. These FGD by-products are often alkaline and contain many plant nutrients. Land application of FGD by-products is encouraged but little information is available related to plant responses and environmental impacts concerning such use. Agricultural lime (ag-lime) and several new types of

L Chen; W. A Dick; S Nelson

2001-01-01

381

Metal Production in Quasars Through Jet-Gas Interactions  

NASA Astrophysics Data System (ADS)

Emission lines studies of the gas surrounding many high redshift quasars indicate a high concentration of CNO nuclides. Relative abundance ratios may even exceed solar levels in some objects with redshifts near 5.0, indicating a rapid buildup of metals within one billion years after the big bang. Models explaining these high concentrations through standard stellar processing are pressed by the short time requirement. We explore a non-stellar nucleosynthesis mechanism in quasars based on the interaction of a high energy particle jet with hot, relatively dense gas. Although temperatures in the hot gas are high enough to support (thermalized) thermonuclear reactions, this mechanism alone is too slow to allow a rapid buildup of CNO nuclides. The collision of (non-thermal) jet particles with gas particles allows creation of unique nuclides which can boost the nucleosynthesis over traditional mass gaps at A = 5 and A = 8. The temperature and initial particle density range from T9=0.2 to T9=5.0, and 1011 to 1018 particles/cm3, respectively, while the jet intensity varies from 0.1 to 10 solar masses per year. The maximum final density allowed is 1023 particles/cm3. Substantial metal production in just 100 days can occur for temperatures near T9=0.6 and final densities of 1021 particles/cm3. Production at other temperatures and densities varies greatly. If the temperature is much above or below T9=0.6, or if the density cannot reach 1021 particles/cm3, then metal production is limited. Although the simple jet-clump model by itself does not seem capable of fully explaining the solar abundances in quasar gas, the low level production occurs on sufficiently short time scales so that it is still interesting. Also, a simplistic exploration of the production resulting from gas which evolves from high to low temperatures seems to indicate that at least 1/100th of solar levels can be obtained if the density can climb to 1021 particles/cm3 in a single processing episode of about 200 days. Multiple processing episodes and more complicated cooling scenarios may indicate larger nucleosynthesis possibilities. Therefore, the jet-clump model offers an exciting possibility for generating metals in quasars.

Vandegriff, Jon D.

382

Natural biodegradation may suffice at natural gas production sites  

SciTech Connect

Biodegradation of organic contaminants in soil and ground water by indigenous microbes is very common, but the contribution of this intrinsic process to site remediation has sometimes been overlooked as too slow or ineffective. Some recent RCRA corrective measure studies and CERCLA remedial investigation/feasibility studies are challenging this mentality. One recent investigation evaluated the potential for intrinsic bioremediation to act as the sole remedial measure at a natural gas production site, possibly supporting a no-intervention decision. The site in question is contaminated with hydrocarbon condensate from natural gas production wells. Amoco Oil (Tulsa, Oklahoma), which operates the wells, evaluated intrinsic aerobic and anaerobic bioremediation at that location. Preliminary results are promising. This evaluation is described in this paper. 1 ref., 1 fig., 1 tab.

NONE

1996-05-01

383

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

Code of Federal Regulations, 2010 CFR

...rigid gas permeable contact lens care product is a device...storing of a rigid gas permeable contact lens. This includes all solutions and tablets used together with rigid gas permeable contact lenses. (b)...

2010-04-01

384

Production of bioplastics and hydrogen gas by photosynthetic microorganisms  

Microsoft Academic Search

Our efforts have been aimed at the technological basis of photosynthetic-microbial production of materials and an energy carrier.\\u000a We report here accumulation of poly-(3-hydroxybutyrate) (PHB), a raw material of biodegradable plastics and for production\\u000a of hydrogen gas, and a renewable energy carrier by photosynthetic microorganisms (tentatively defined as cyanobacteria plus\\u000a photosynthetic bateria, in this report).\\u000a \\u000a A thermophilic cyanobacterium,Synechococcus sp. MA19

Asada Yasuo; Miyake Masato; Miyake Jun

1998-01-01

385

Mineralogical and engineering characteristics of dry flue gas desulfurization products  

Microsoft Academic Search

Fifty-nine coal combustion products were collected from coal-fired power plants using various dry flue gas desulfurization (FGD) processes to remove SO2. X-ray diffraction analyses revealed duct injection and spray dryer processes created products that primarily contained Ca(OH)2 (portlandite) and CaSO3·0.5H2O (hannebachite). Most samples from the lime injection multistage burners process contained significant amounts of CaO (lime), CaSO4 (anhydrite), and CaCO3

Jerry M. Bigham; David A. Kost; Richard C. Stehouwer; Joel H. Beeghly; Randy Fowler; Samuel J. Traina; William E. Wolfe; Warren A. Dick

2005-01-01

386

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; Schlüter, Stefan; Hennig, Torsten; Deerberg, Görge

2013-04-01

387

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

NASA Astrophysics Data System (ADS)

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

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

2014-03-01

388

MANTRA: An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra  

NASA Astrophysics Data System (ADS)

This paper presents an update of an on-going collaborative INL-ANL-ISU integral reactor physics experiment whose objective is to infer the effective neutron capture cross sections for most of the actinides of importance for reactor physics and fuel cycle studies in both fast and epithermal spectra. Some fission products are also being considered. The principle of the experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation together with the neutron fluence will allow inference of effective neutron capture cross-sections in different neutron spectra.

Youinou, G.; Vondrasek, R.; Veselka, H.; Salvatores, M.; Paul, M.; Pardo, R.; Palmiotti, G.; Palchan, T.; Nusair, O.; Nimmagadda, J.; Nair, C.; Murray, P.; Maddock, T.; Kondrashev, S.; Kondev, F. G.; Jones, W.; Imel, G.; Glass, C.; Fonnesbeck, J.; Berg, J.; Bauder, W.

2014-05-01

389

Process for the production of fuel gas from coal  

DOEpatents

An improved apparatus and process for the conversion of hydrocarbonaceous materials, such as coal, to more valuable gaseous products in a fluidized bed gasification reaction and efficient withdrawal of agglomerated ash from the fluidized bed is disclosed. The improvements are obtained by introducing an oxygen containing gas into the bottom of the fluidized bed through a separate conduit positioned within the center of a nozzle adapted to agglomerate and withdraw the ash from the bottom of the fluidized bed. The conduit extends above the constricted center portion of the nozzle and preferably terminates within and does not extend from the nozzle. In addition to improving ash agglomeration and withdrawal, the present invention prevents sintering and clinkering of the ash in the fluidized bed and permits the efficient recycle of fine material recovered from the product gases by contacting the fines in the fluidized bed with the oxygen as it emanates from the conduit positioned within the withdrawal nozzle. Finally, the present method of oxygen introduction permits the efficient recycle of a portion of the product gases to the reaction zone to increase the reducing properties of the hot product gas.

Patel, Jitendra G. (Bolingbrook, IL); Sandstrom, William A. (Chicago, IL); Tarman, Paul B. (Elmhurst, IL)

1982-01-01

390

Multiphasic analysis of gas production kinetics for in vitro fermentation of ruminant feeds  

Microsoft Academic Search

Recently developed time-related gas production techniques to quantify the kinetics of ruminant feed fermentation have a high resolution. Consequently, fermentation processes with clearly contrasting gas production kinetics can be identified. Parameterization of the separate processes is possible with a suitable multiphasic model and modelling method. A flexible, empirical, multiphasic model was proposed for parameterization of gas production profiles. This equation

Jeroen C. J. Groot; John W. Cone; Barbara A. Williams; Filip M. A. Debersaques; Egbert A. Lantinga

1996-01-01

391

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

NASA Astrophysics Data System (ADS)

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

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

1985-05-01

392

Production of biofuels from synthesis gas using microbial catalysts.  

PubMed

World energy consumption is expected to increase 44% in the next 20 years. Today, the main sources of energy are oil, coal, and natural gas, all fossil fuels. These fuels are unsustainable and contribute to environmental pollution. Biofuels are a promising source of sustainable energy. Feedstocks for biofuels used today such as grain starch are expensive and compete with food markets. Lignocellulosic biomass is abundant and readily available from a variety of sources, for example, energy crops and agricultural/industrial waste. Conversion of these materials to biofuels by microorganisms through direct hydrolysis and fermentation can be challenging. Alternatively, biomass can be converted to synthesis gas through gasification and transformed to fuels using chemical catalysts. Chemical conversion of synthesis gas components can be expensive and highly susceptible to catalyst poisoning, limiting biofuel yields. However, there are microorganisms that can convert the CO, H(2), and CO(2) in synthesis gas to fuels such as ethanol, butanol, and hydrogen. Biomass gasification-biosynthesis processing systems have shown promise as some companies have already been exploiting capable organisms for commercial purposes. The discovery of novel organisms capable of higher product yield, as well as metabolic engineering of existing microbial catalysts, makes this technology a viable option for reducing our dependency on fossil fuels. PMID:20359454

Tirado-Acevedo, Oscar; Chinn, Mari S; Grunden, Amy M

2010-01-01

393

75 FR 20271 - Oil and Gas and Sulphur Operations in the Outer Continental Shelf-Oil and Gas Production...  

Federal Register 2010, 2011, 2012, 2013

...by the cost recovery program at budget for MMS in...will commit to the development of a information to...quantities within the oil and gas these into our regulations...reporting is required for developments in the Alaska OCS Region...the requirements for oil and gas production...produce the oil and gas at the rates that......

2010-04-19

394

[Determination of thiocyanate in dairy products by headspace gas chromatography].  

PubMed

A method for the determination of thiocyanate in dairy products by headspace gas chromatography was established. At first, the thiocyanate in dairy products was extracted by water. Then, the zinc acetate solution was added to the crude product for protein precipitation. The extract obtained above was centrifuged and the supernatant was added with chloramine T, which derivatized the thiocyanate ions to cyanogen chloride. The head-space vapor of the final extract was injected into a BP10 (14% cyanopropyl phenyl polysiloxane) gas chromatographic column, and detected by an electron capture detector (ECD). The target compound was quantified by external standard. The results showed that there was a good linearity between 0.005 mg/L and 0.1 mg/L with the correlation coefficient (r) of 0.997, and the limit of detection (signal-to-noise ratio (S/N) > or = 10) was 0.1 mg/kg. The recoveries were 90.0%-110.0% with the relative standard deviations (RSDs) (n = 10) of 4.98%-7.89% at the three spiked levels of 1.0, 2.0, 10.0 mg/kg. In conclusion, this method is simple, rapid and accurate. It can be applied in the determination of thiocyanate in dairy products, and meets the requirements of the daily testing. The method has been successfully used to test 18 kinds of commercially available dairy products and it was found that all the dairy products tested contained thiocyanate about 0.5-10 mg/kg PMID:23189673

Song, Jie; Fu, Yingwen; Du, Lijun; Yao, Yating; Mu, Xiajie; Kang, Lihua; Zhao, Youyou

2012-07-01

395

Production of manufactured aggregates from flue gas desulfurization by-products  

Microsoft Academic Search

CONSOL R and D has developed a disk pelletization process to produce manufactured aggregates from the by-products of various technologies designed to reduce sulfur emissions produced from coal utilization. Aggregates have been produced from the by-products of the Coolside and LIMB sorbent injection, the fluidized-bed combustion (FBC), spray dryer absorption (SDA), and lime and limestone wet flue gas desulfurization (FGD)

M. M. Wu; D. C. McCoy; M. L. Fenger; R. O. Scandrol; R. A. Winschel; J. A. Withum; R. M. Statnick

1999-01-01

396

Characterizing tight-gas systems with production data: Wyoming, Utah, and Colorado  

USGS Publications Warehouse

The study of produced fluids allows comparisons among tight-gas systems. This paper examines gas, oil, and water production data from vertical wells in 23 fields in five Rocky Mountain basins of the United States, mostly from wells completed before the year 2000. Average daily rates of gas, oil, and water production are determined two years and seven years after production begins in order to represent the interval in which gas production declines exponentially. In addition to the daily rates, results are also presented in terms of oil-to-gas and water-to-gas ratios, and in terms of the five-year decline in gas production rates and water-to-gas ratios. No attempt has been made to estimate the ultimate productivity of wells or fields. The ratio of gas production rates after seven years to gas production rates at two years is about one-half, with median ratios falling within a range of 0.4 to 0.6 in 16 fields. Oil-gas ratios show substantial variation among fields, ranging from dry gas (no oil) to wet gas to retrograde conditions. Among wells within fields, the oil-gas ratios vary by a factor of three to thirty, with the exception of the Lance Formation in Jonah and Pinedale fields, where the oil-gas ratios vary by less than a factor of two. One field produces water-free gas and a large fraction of wells in two other fields produce water-free gas, but most fields have water-gas ratios greater than 1 bbl/mmcf—greater than can be attributed to water dissolved in gas in the reservoir— and as high as 100 bbl/mmcf. The median water-gas ratio for fields increases moderately with time, but in individual wells water influx relative to gas is erratic, increasing greatly with time in many wells while remaining constant or decreasing in others.

Nelson, Philip H.; Santus, Stephen L.

2013-01-01

397

Trash to Gas: Converting Space Waste into Useful Supply Products  

NASA Technical Reports Server (NTRS)

The cost of sending mass into space with current propulsion technology is very expensive, making every item a crucial element of the space mission. It is essential that all materials be used to their fullest potential. Items like food, packaging, clothing, paper towels, gloves, etc., normally become trash and take up space after use. These waste materials are currently either burned up upon reentry in earth's atmosphere or sent on cargo return vehicles back to earth: a very wasteful method. The purpose of this project was to utilize these materials and create useful products like water and methane gas, which is used for rocket fuel, to further supply a deep space mission. The system used was a thermal degradation reactor with the configuration of a down-draft gasifier. The reactor was loaded with approximately 100g of trash simulant and heated with two external ceramic heaters with separate temperature control in order to create pyrolysis and gasification in one zone and incineration iri a second zone simultaneously. Trash was loaded into the top half of the reactor to undergo pyrolysis while the downdraft gas experienced gasification or incineration to treat tars and maximize the production of carbon dioxide. Minor products included carbon monoxide, methane, and other hydrocarbons. The carbon dioxide produced can be sent to a Sabatier reactor to convert the gas into methane, which can be used as rocket propellant. In order to maximize the carbon dioxide and useful gases produced, and minimize the unwanted tars and leftover ashen material, multiple experiments were performed with altered parameters such as differing temperatures, flow rates, and location of inlet air flow. According to the data received from these experiments, the process will be further scaled up and optimized to ultimately create a system that reduces trash buildup while at the same time providing enough useful gases to potentially fill a methane tank that could fuel a lunar ascent vehicle or other deep space mission.

Tsoras, Alexandra

2013-01-01

398

Vitrification of fission product solutions: Investigation of the effects of noble metals on the fabrication and properties of R7T7 glass  

SciTech Connect

The effects of incorporating additional qualities of insoluble dissolution fines containing noble metals together with the fission product feed solutions were investigated for application to the French T7 vitrification facility at La Hague. Three types of tests were conducted: nonradioactive laboratory tests, radioactive laboratory tests, and industrial-scale tests in a prototype vitrification unit. The laboratory test results showed that the quality of R7T7 glass containing from 1.5 to 4 wt% of platinoids is fully equivalent to that of standard R7T7 glass without platinoids. These findings were confirmed on glass casting samples containing 0 to 3 wt% of platinoids from a full-scale industrial vitrification prototype facility. Recent tests in which video cameras have been used to visualize the molten glass and model simulations of glass properties and of melting pot behavior suggest that industrial operating conditions can be optimized to produce glass with platinoid concentrations approaching this limit value.

Puyou, M.; Jacquet-Francillon, N.; Moncouyoux, J.P.; Sombret, C.; Teulon, F. [Commissariat a l`Energie Atomique, Bagnols-sur-Ceze (France)

1995-07-01

399

Fission-suppressed blankets for fissile fuel breeding fusion reactors  

Microsoft Academic Search

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions.

J. D. Lee; R. W. Moir

1981-01-01

400

Large-scale shell model calculations for odd-odd nuclei and comparison to experimental studies of fission product nuclei in the /sup 132/Sn region  

SciTech Connect

Experimental spectroscopy data of fission products have been obtained using highly automated and rapid chemical separations followed by automated spectroscopy studies of isolated fission products. These data have established the presence of only a single level with spin-parity of 1/sup +/ below 1500 keV of excitation in Z = 51 /sup 132/Sb/sub 81/. This is in contrast to the results of our studies of /sup 130/Sb and /sup 134/I. For /sup 134/I, the N = 81 isotone with Z = 53, we can characterize three 1/sup +/ levels below 1200 keV. For /sup 130/Sb/sub 79/ that has a neutron pair less than /sup 132/Sb, we can identify two 1/sup +/ levels below 1100 keV. We can account for the additional levels using the LLNL shell-model code which is based on the Lanczos tridiagonalization algorithm using an uncoupled m-scheme basis and vector manipulations. The 1g/sub 7/2/, 2d/sub 5/2/, 2d/sub 3/2/, 1h/sub 11/2/, and 3s/sub 1/2/ orbitals are available to the valence protons and the 2d/sub 5/2/, 2d/sub 3/2/, 1h/sub 11/2/, and 3s/sub 1/2/ orbitals are available to the valence neutron holes. Analysis of the wavefunctions show the dominant role of three nucleon cluster configurations in producing the increased number of states at low energy. The absence of nucleon cluster configurations in the parent nucleus /sup 130/Sn is used to explain the reduction of approximately a factor of 20 in the Gamow-Teller beta strength to the low lying 1/sup +/ levels of /sup 130/Sb. 27 references.

Lane, S.M.; Henry, E.A.; Meyer, R.A.

1985-01-08

401

Air quality concerns of unconventional oil and natural gas production.  

PubMed

Increased use of hydraulic fracturing ("fracking") in unconventional oil and natural gas (O & NG) development from coal, sandstone, and shale deposits in the United States (US) has created environmental concerns over water and air quality impacts. In this perspective we focus on how the production of unconventional O & NG affects air quality. We pay particular attention to shale gas as this type of development has transformed natural gas production in the US and is set to become important in the rest of the world. A variety of potential emission sources can be spread over tens of thousands of acres of a production area and this complicates assessment of local and regional air quality impacts. We outline upstream activities including drilling, completion and production. After contrasting the context for development activities in the US and Europe we explore the use of inventories for determining air emissions. Location and scale of analysis is important, as O & NG production emissions in some US basins account for nearly 100% of the pollution burden, whereas in other basins these activities make up less than 10% of total air emissions. While emission inventories are beneficial to quantifying air emissions from a particular source category, they do have limitations when determining air quality impacts from a large area. Air monitoring is essential, not only to validate inventories, but also to measure impacts. We describe the use of measurements, including ground-based mobile monitoring, network stations, airborne, and satellite platforms for measuring air quality impacts. We identify nitrogen oxides, volatile organic compounds (VOC), ozone, hazardous air pollutants (HAP), and methane as pollutants of concern related to O & NG activities. These pollutants can contribute to air quality concerns and they may be regulated in ambient air, due to human health or climate forcing concerns. Close to well pads, emissions are concentrated and exposure to a wide range of pollutants is possible. Public health protection is improved when emissions are controlled and facilities are located away from where people live. Based on lessons learned in the US we outline an approach for future unconventional O & NG development that includes regulation, assessment and monitoring. PMID:24699994

Field, R A; Soltis, J; Murphy, S

2014-05-01

402

Greenhouse gas emission associated with sugar production in southern Brazil  

PubMed Central

Background Since sugarcane areas have increased rapidly in Brazil, the contribution of the sugarcane production, and, especially, of the sugarcane harvest system to the greenhouse gas emissions of the country is an issue of national concern. Here we analyze some data characterizing various activities of two sugarcane mills during the harvest period of 2006-2007 and quantify the carbon footprint of sugar production. Results According to our calculations, 241 kg of carbon dioxide equivalent were released to the atmosphere per a ton of sugar produced (2406 kg of carbon dioxide equivalent per a hectare of the cropped area, and 26.5 kg of carbon dioxide equivalent per a ton of sugarcane processed). The major part of the total emission (44%) resulted from residues burning; about 20% resulted from the use of synthetic fertilizers, and about 18% from fossil fuel combustion. Conclusions The results of this study suggest that the most important reduction in greenhouse gas emissions from sugarcane areas could be achieved by switching to a green harvest system, that is, to harvesting without burning. PMID:20565736

2010-01-01

403

Storage sizing for embedding of local gas production in a micro gas grid  

NASA Astrophysics Data System (ADS)

In this paper we study the optimal control of a micro grid of biogas producers. The paper considers the possibility to have a local storage device for each producer, who partly consumes his own production, i.e. prosumer. In addition, connected prosumers can sell stored gas to create revenue from it. An optimization model is employed to derive the size of storage device and to provide a pricing mechanism in an effort to value the stored gas. Taking into account physical grid constraints, the model is constructed in a centralized scheme of model predictive control. Case studies show that there is a relation between the demand and price profiles in terms of peaks and lows. The price profiles generally follow each other. The case studies are employed as well to to study the impacts of model parameters on deriving the storage size.

Alkano, D.; Nefkens, W. J.; Scherpen, J. M. A.; Volkerts, M.

2014-12-01

404

Producing ammonium sulfate from flue gas desulfurization by-products  

USGS Publications Warehouse

Emission control technologies using flue gas desulfurization (FGD) have been widely adopted by utilities burning high-sulfur fuels. However, these technologies require additional equipment, greater operating expenses, and increased costs for landfill disposal of the solid by-products produced. The financial burdens would be reduced if successful high-volume commercial applications of the FGD solid by-products were developed. In this study, the technical feasibility of producing ammonium sulfate from FGD residues by allowing it to react with ammonium carbonate in an aqueous solution was preliminarily assessed. Reaction temperatures of 60, 70, and 80??C and residence times of 4 and 6 hours were tested to determine the optimal conversion condition and final product evaluations. High yields (up to 83%) of ammonium sulfate with up to 99% purity were achieved under relatively mild conditions. The optimal conversion condition was observed at 60??C and a 4-hour residence time. The results of this study indicate the technical feasibility of producing ammonium sulfate fertilizer from an FGD by-product. Copyright ?? Taylor & Francis Inc.

Chou, I.-M.; Bruinius, J.A.; Benig, V.; Chou, S.-F.J.; Carty, R.H.

2005-01-01

405

Notes on Sable Natural Gas Production December 1999 to November 2005  

E-print Network

, as well as its embracing of the proposed liquefied natural gas (LNG) facilities at Bear Head and GuysboroERG/200601 Notes on Sable Natural Gas Production December 1999 to November 2005 Larry Hughes Energy;Hughes: Notes on Sable Natural Gas Production 1 1. Background The Sable Offshore Energy Project consists

Hughes, Larry

406

Total greenhouse gas emissions related to the Dutch crop production system  

Microsoft Academic Search

This article discusses the greenhouse gas emissions (CO2, CH4, N2O) related to Dutch agricultural crop production. Emissions occur during agricultural processes (direct emissions) as well as in the life cycle of the required inputs (indirect emissions). An integrated approach assesses the total greenhouse gas emissions related to Dutch agricultural crop production. The results show differences in total greenhouse gas emissions

Henri C. Moll; Sanderine Nonhebel

1999-01-01

407

Natural gas productive capacity for the lower 48 states 1984 through 1996, February 1996  

SciTech Connect

This is the fourth wellhead productive capacity report. The three previous ones were published in 1991, 1993, and 1994. This report should be of particular interest to those in Congress, Federal and State agencies, industry, and the academic community, who are concerned with the future availability of natural gas. The EIA Dallas Field Office has prepared five earlier reports regarding natural gas productive capacity. These reports, Gas Deliverability and Flow Capacity of Surveillance Fields, reported deliverability and capacity data for selected gas fields in major gas producing areas. The data in the reports were based on gas-well back-pressure tests and estimates of gas-in-place for each field or reservoir. These reports use proven well testing theory, most of which has been employed by industry since 1936 when the Bureau of Mines first published Monograph 7. Demand for natural gas in the United States is met by a combination of natural gas production, underground gas storage, imported gas, and supplemental gaseous fuels. Natural gas production requirements in the lower 48 States have been increasing during the last few years while drilling has remained at low levels. This has raised some concern about the adequacy of future gas supplies, especially in periods of peak heating or cooling demand. The purpose of this report is to address these concerns by presenting a 3-year projection of the total productive capacity of natural gas at the wellhead for the lower 48 States. Alaska is excluded because Alaskan gas does not enter the lower-48 States pipeline system. The Energy Information Administration (EIA) generates this 3-year projection based on historical gas-well drilling and production data from State, Federal, and private sources. In addition to conventional gas-well gas, coalbed gas and oil-well gas are also included.

NONE

1996-02-09

408

Ballistic piston fissioning plasma experiment.  

NASA Technical Reports Server (NTRS)

The production of fissioning uranium plasma samples such that the fission fragment stopping distance is less than the dimensions of the plasma is approached by using a ballistic piston device for the compression of uranium hexafluoride. The experimental apparatus is described. At room temperature the gun can be loaded up to 100 torr UF6 partial pressure, but at compression a thousand fold increase of pressure can be obtained at a particle density on the order of 10 to the 19th power per cu cm. Limited spectral studies of UF6 were performed while obtaining the pressure-volume data. The results obtained and their implications are discussed.

Miller, B. E.; Schneider, R. T.; Thom, K.; Lalos, G. T.

1971-01-01

409

Ground movements associated with gas hydrate production. Final report  

SciTech Connect

This report deals with a study directed towards a modeling effort on production related ground movements and subsidence resulting from hydrate dissociation. The goal of this research study was to evaluate whether there could be subsidence related problems that could be an impediment to hydrate production. During the production of gas from a hydrate reservoir, it is expected that porous reservoir matrix becomes more compressible which may cause reservoir compression (compaction) under the influence of overburden weight. The overburden deformations can propagate its influence upwards causing subsidence near the surface where production equipment will be located. In the present study, the reservoir compaction is modeled by using the conventional ``stress equilibrium`` approach. In this approach, the overburden strata move under the influence of body force (i.e. self weight) in response to the ``cavity`` generated by reservoir depletion. The present study is expected to provide a ``lower bound`` solution to the subsidence caused by hydrate reservoir depletion. The reservoir compaction anticipated during hydrate production was modeled by using the finite element method, which is a powerful computer modeling technique. The ground movements at the reservoir roof (i.e. reservoir compression) cause additional stresses and disturbance in the overburden strata. In this study, the reservoir compaction was modeled by using the conventional ``stress equilibrium`` approach. In this approach, the overburden strata move under the influence of body force (i.e. self weight) in response to the ``cavity`` generated by reservoir depletion. The resulting stresses and ground movements were computed by using the finite element method. Based on the parameters used in this investigation, the maximum ground subsidence could vary anywhere from 0.50 to 6.50 inches depending on the overburden depth and the size of the depleted hydrate reservoir.

Siriwardane, H.J.; Kutuk, B.

1992-03-01

410

Thorium-uranium fission radiography  

NASA Technical Reports Server (NTRS)

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

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

1976-01-01

411

Commander field: Case study of a gas productive landsat and geochemical anomaly, Parker County, Texas  

SciTech Connect

Landsat data for a mature area of southern Parker County were analyzed for structural anomalies and lineaments in order to determine the relationship of these data to gas production and the possible fracturing of Atoka sand reservoirs. Interstitial soil gas data gathered over a 160 acre tract revealed a strong surface anomaly situated at the intersection of two lineaments. The drilling of this anomaly resulted in gas production from a Bend conglomerate and excellent mudlog shows of gas from shallower sands in the Atoka and Strawn intervals. A subsequent offset well, located within the original surface soil gas anomaly, also proved gas productive in the shallow Strawn interval. Well data from the productive gas zones are discussed in relation to local stratigraphy and structure. The limitations and advantages of Landsat/soil gas data are considered in terms of future applicability to other mature areas.

Crowder, W.T. Jr. [Consulting Geologist, Dallas, TX (United States)

1995-06-01

412

Atmospheric emissions and air quality impacts from natural gas production and use.  

PubMed

The US Energy Information Administration projects that hydraulic fracturing of shale formations will become a dominant source of domestic natural gas supply over the next several decades, transforming the energy landscape in the United States. However, the environmental impacts associated with fracking for shale gas have made it controversial. This review examines emissions and impacts of air pollutants associated with shale gas production and use. Emissions and impacts of greenhouse gases, photochemically active air pollutants, and toxic air pollutants are described. In addition to the direct atmospheric impacts of expanded natural gas production, indirect effects are also described. Widespread availability of shale gas can drive down natural gas prices, which, in turn, can impact the use patterns for natural gas. Natural gas production and use in electricity generation are used as a case study for examining these indirect consequences of expanded natural gas availability. PMID:24498952

Allen, David T

2014-01-01

413

Prediction of Gas Leak Tightness of Superplastically Formed Products  

SciTech Connect

In some applications, in this case an aluminium box in a subatomic particle detector containing highly sensitive detecting devices, it is important that a formed sheet should show no gas leak from one side to the other. In order to prevent a trial-and-error procedure to make this leak tight box, a method is set up to predict if a formed sheet conforms to the maximum leak constraint. The technique of superplastic forming (SPF) is used in order to attain very high plastic strains before failure. Since only a few of these boxes are needed, this makes, this generally slow, process an attractive production method. To predict the gas leak of a superplastically formed aluminium sheet in an accurate way, finite element simulations are used in combination with a user-defined material model. This constitutive model couples the leak rate with the void volume fraction. This void volume fraction is then dependent on both the equivalent plastic strain and the applied hydrostatic pressure during the bulge process (backpressure).

Snippe, Corijn H. C. [National Institute for Subatomic Physics (Nikhef) PO Box 41882, 1009 DB Amsterdam (Netherlands); Meinders, T. [University of Twente, Faculty of Engineering Technology PO Box 217, 7500 AE Enschede (Netherlands)

2010-06-15

414

Life-Cycle Greenhouse Gas and Energy Analyses of Algae Biofuels Production  

E-print Network

Life-Cycle Greenhouse Gas and Energy Analyses of Algae Biofuels Production Transportation Energy algae life cycle studies and compare new analyses of multiple production scenarios and process the life cycle greenhouse gas emissions, net energy consumption, and net liquid fuels production

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