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1

Dosimetry of noble-gas fission products  

Microsoft Academic Search

In the first minutes and hours following a reactor accident, personnel at the reactor facility, and possibly members of the public at off-site locations may be subjected to significant radiation dose from exposure to the entire spectrum of noble gas fission products and their daughter radionuclides. In order to measure the immersion dose of noble gas fission products following the

P. J. T

1989-01-01

2

Hydrogen production from nuclear fission product waste heat and use in gas turbines  

Microsoft Academic Search

An analysis has been made on the feasibility of producing hydrogen using fission product waste heat and its subsequent combustion in gas turbines. The work has been performed in three distinct phases. In the first phase, a system using radioactive waste heat has been designed, which produces electricity. The electrical power output of this system has been calculated as a

M. E. Nelson; E. L. Keating; D. R. Govan; R. J. Banchak; J. R. Corpus

1979-01-01

3

New Results on Helium and Tritium Gas Production From Ternary Fission  

NASA Astrophysics Data System (ADS)

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.; Wagemans, C.; Heyse, J.

2005-05-01

4

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

NASA Astrophysics Data System (ADS)

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

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

2010-02-01

5

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

6

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

SciTech Connect

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

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

2008-09-01

7

Payload Dose Rate from Direct Beam Radiation and Exhaust Gas Fission Products.  

National Technical Information Service (NTIS)

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 th...

M. A. Capo R. Mickle

1975-01-01

8

Morphology of fission gas bubbles in fissioning uranium metal closely  

NASA Astrophysics Data System (ADS)

We investigate by SEM the micro-structural and basic phenomenological mechanisms governing the fission-gas and fusion-gas behaviour in metals. This comparative study clearly shows the characteristics of fission-gas bubbles (primarily helium and xenon) in uranium fuel metals have the same characteristics as fusion-gas bubbles (helium) in the solid-state fusion metal - palladium. The remarkably similar characteristic morphology clearly identifies the nuclear phenomenological origins of the gas bubbles in the palladium metal which are correllated and explained by the presence of a large amount of DD fusion. Allied evidence of anomalous heat production during cold fusion experiments suggests the nuclear process. Further analysis of these fusion metals by mass spectroscopy clearly identifies anomalous helium isotopes in large quantity were trapped in the palladium metal.

George, Russ

2005-03-01

9

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

SciTech Connect

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

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

1987-01-01

10

FISSION PRODUCT GAMMA RAY SPECTRA  

Microsoft Academic Search

Uranium-235 fission product gamma spectra were calculated for various ; reactor operating histories. An IBM704 digital computer program was coded to ; compute the gamma energy contributed by each fission product gamma ray and to sum ; these results in energy groups. A representative curve showing the decay of ; various energy groups is presented. Comparisons are made with data

1958-01-01

11

TREATMENT OF FISSION PRODUCT WASTE  

DOEpatents

A pyrogenic method of separating nuclear reactor waste solutions containing aluminum and fission products as buring petroleum coke in an underground retort, collecting the easily volatile gases resulting as the first fraction, he uminum chloride as the second fraction, permitting the coke bed to cool and ll contain all the longest lived radioactive fission products in greatly reduced volume.

Huff, J.B.

1959-07-28

12

Evaluation of Fission Product After-Heat.  

National Technical Information Service (NTIS)

Reported here are studies on: estimation of fission-product gamma spectra; comparisons of theoretical predictions of fission-product decay power with ongoing experimental programs; and a protocol for estimating bias in decay energy estimates for fission p...

B. I. Spinrad

1976-01-01

13

REMOVAL OF FISSION PRODUCT GASES FROM REACTOR OFF-GAS STREAMS BY ADSORPTION (PRESENTED AT AMERICAN NUCLEAR SOCIETY MEETING, DETROIT, MICHIGAN, DECEMBER 10, 1958)  

Microsoft Academic Search

A disposal process is described in which the noble gas fission products, ;\\u000a krypton and xenon, are delayed relative to the sweep gas by physical adsorption ;\\u000a as they pass through an adsorbent such as activated charcoal. A theoretical ;\\u000a expression describing this process was developed, using a theoretical plate ;\\u000a analysis, and was verified experimentally. The retention time for

W. E. Browning; R. E. Adams; R. D. Ackley

1959-01-01

14

Correlation of recent fission product release data  

SciTech Connect

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

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

1989-01-01

15

Evaluation of Fission Product Afterheat.  

National Technical Information Service (NTIS)

This report covers work under the subject contract during the spring quarter of 1977. The goal of the contract is to improve the understanding of shutdown power in reactors due to radioactive decay of fission products, particularly for light water reactor...

B. I. Spinrad

1977-01-01

16

Behavior of solid fission products in coated fuel particles of a high-temperature gas-cooled reactor  

Microsoft Academic Search

To study the retention properties of protective coatings, we conducted an experimental study of the profiles of the volume concentrations of nuclear fuel and its fission products in sample HTGR coated fuel particles (of the TRISO type) that have been through in-pile testing over a wide range of temperatures (1273-2133 K) and burnup (0.1-17% fima). To provide a more reliable

A. N. Gudkov; V. A. Kashparov; A. A. Kotlyarov; N. N. Ponomarev-Stepnoi; I. G. Prikhod'ko; A. A. Khrulev

1989-01-01

17

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

NASA Astrophysics Data System (ADS)

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

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

1996-10-01

18

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

SciTech Connect

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

John K. Hartwell

2007-06-01

19

RECOVERY OF ALUMINUM FROM FISSION PRODUCTS  

DOEpatents

A method is given for recovertng aluminum values from aqueous solutions containing said values together with fission products. A mixture of Fe/sub 2/O/ sub 3/ and MnO/sub 2/ is added to a solution containing aluminum and fission products. The resulting aluminum-containing supernatant is then separated from the fission product-bearing metal oxide precipitate and is contacted with a cation exchange resin. The aluminum sorbed on the resin is then eluted and recovered. (AEC)

Blanco, R.E.; Higgins, I.R.

1962-11-20

20

METHOD FOR SEPARATING PLUTONIUM AND FISSION PRODUCTS EMPLOYING AN OXIDE AS A CARRIER FOR FISSION PRODUCTS  

DOEpatents

Carrier precipitation processes for separating plutonium values from uranium fission products are described. Silicon dioxide or titanium dioxide in a finely divided state is added to an acidic aqueous solution containing hexavalent plutonium ions together with ions of uranium fission products. The supernatant solution containing plutonium ions is then separated from the oxide and the fission products associated therewith.

Davies, T.H.

1961-07-18

21

Nondestructive fission gas release measurement and analysis  

Microsoft Academic Search

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

P. M. OLeary; D. R. Packard

1993-01-01

22

Rapid separation of fresh fission products (draft)  

Microsoft Academic Search

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

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

2003-01-01

23

Thermochromatographic Investigations of Fission Product Transport and Chemistry.  

National Technical Information Service (NTIS)

A thermochromatographic technique has been developed to investigate the chemical states of fission products from irradiated fuel as well as in fission product simulation studies. Some recent work on iodine transport and on release of fission products from...

F. B. Growcock S. Aronson M. Friedlander J. Skalyo A. Hosseini

1978-01-01

24

TMI-2 Fission Product Inventory Estimates (Draft).  

National Technical Information Service (NTIS)

This report presents the results of analyses performed to estimate the inventory and distribution of selected radioisotopes within the TMI-2 reactor system. The intent of the report is to document the method used in estimating the fission product inventor...

E. L. Tolman M. Nishio

1987-01-01

25

Formation and transport of fission product aerosols  

SciTech Connect

The calculation of fission product retention is the primary system of light water reactors requires the knowledge of the size distribution of the particles formed from the condensation of the volatile fission product vapors released in LWR accidents. To this end, a computer model RAFT (Reactor Aerosol Formation and Transport) is being developed to predict the size distribution and composition of the condensed cesium, iodine and tellurium containing aerosols.

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

1983-01-01

26

Transport properties of fission product vapors  

SciTech Connect

Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors.

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

1983-07-01

27

ORNL fission product release tests VI6  

Microsoft Academic Search

The ORNL fission product release tests investigate release and transport of the major fission products from high-burnup fuel under LWR accident conditions. The two most recent tests (VI-4 and VI-5) were conducted in hydrogen. In three previous tests in this series (VI-1, VI-2, and VI-3), which had been conducted in steam, the oxidized Zircaloy cladding remained largely intact and acted

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

1991-01-01

28

Systematics of Fission-Product Yields  

SciTech Connect

Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

A.C. Wahl

2002-05-01

29

A NEW FISSION PRODUCT: Ga⁷⁴  

Microsoft Academic Search

The 8 min Ga⁷⁴ was separated from the fission-products of U\\/sup ; 235\\/. The gallium activity was characterized as Ga⁷⁴ by comparison of its ; gamma -ray spectrum with that obtained from (n,p)-produced Ga⁷⁴. Chemical ; separation of the Ga⁷⁴ was effected by a combination of extraction, ; volatilization, ionexchange, and precipitation steps. Its fission yield was ; determined to

J. A. Marinsky; E. Eichler

1960-01-01

30

Fission gas bubbles in uranium-aluminide fuels  

SciTech Connect

Formation of fission gas bubbles heretofore has not been observed in uranium-aluminide fuels. Recent irradiations to record high burnups offered a possibility to determine the onset of fission gas bubble formation in this type of fuel. Present experimental evidence suggests that UAl/sub 2/, UAl/sub 3/, and UAl/sub 4/ do not form fission gas bubbles at fission densities of 7 x 10/sup 21//cm/sup 3/ of fuel (60% depletion of 93% enriched /sup 235/U), and that pure uranium aluminide is likely to remain free of fission gas bubbles to very high /sup 235/U burnup at any enrichment. However, fission gas bubbles were found in these experimental fuels for the first time, but they were without exception associated with uranium-oxide inclusions that were evidently formed during fuel fabrication.

Hofman, G.L.

1987-04-01

31

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

32

Average neutronic properties of prompt fission products  

SciTech Connect

Calculations of the average neutronic properties of the ensemble of fission products producted by fast-neutron fission of /sup 235/U and /sup 239/Pu, where the properties are determined before the first beta decay of any of the fragments, are described. For each case we approximate the ensemble by a weighted average over 10 selected nuclides, whose properties we calculate using nuclear-model parameters deduced from the systematic properties of other isotopes of the same elements as the fission fragments. The calculations were performed primarily with the COMNUC and GNASH statistical-model codes. The results, available in ENDF/B format, include cross sections, angular distributions of neutrons, and spectra of neutrons and photons, for incident-neutron energies between 10/sup -5/ eV and 20 MeV. Over most of this energy range, we find that the capture cross section of /sup 239/Pu fission fragments is systematically a factor of two to five greater than for /sup 235/U fission fragments.

Foster, D.G. Jr.; Arthur, E.D.

1982-02-01

33

Evaluation of Fission Product After-Heat.  

National Technical Information Service (NTIS)

Reported here is work for the period indicated under the subject contract. Specific tasks covered in this report are: (1) Neutron capture in fission products, with the preliminary result that this effect does not lead to a major change in decay power of r...

B. I. Spinrad

1976-01-01

34

TMI2 fission product inventory estimates (draft)  

Microsoft Academic Search

This report presents the results of analyses performed to estimate the inventory and distribution of selected radioisotopes within the TMI-2 reactor system. The intent of the report is to document the method used in estimating the fission product inventory and associated uncertainties. The values presented should be viewed as preliminary. Selected radioisotopes for which best-estimate inventories and uncertainties are presented

E. L. Tolman; M. Nishio

1987-01-01

35

AVR EXPERIMENTS RELATED TO FISSION PRODUCT TRANSPORT  

Microsoft Academic Search

Fission products deposited outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (dba), despite of their relatively small fraction of < 310 -4 of the total inventory. This is, because the maximum tolerable releases in dba are < 10 -7 of the total inventory of

Rainer Moormann

2006-01-01

36

Identification of ?s isomers in fission products  

NASA Astrophysics Data System (ADS)

Several ?s isomers have been observed in the fission products of 241Pu and 239Pu. The detection is based on the time correlation between reaction products selected by the LOHENGRIN spectrometer and the ?-rays depopulating the isomers. New isomers have been observed in 96Rb, 106Nb, 127Sn, and 130Te and the others have been confirmed in 94Y, 126Sn, 129Sn and 129Sb.

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

1998-12-01

37

FISSION GAS HOLDUP TESTS ON HRT CHARCOAL BEDS  

Microsoft Academic Search

Fission gas holdup tests on the HRT charcoal beds under simulated ; operating conditions are complete. A radioactive tracer technique developed for ; use in laboratory absorption studies was utilized. The efficiency of the ; charcoal beds, in regard to holdup of fission gases, exceeds design ; specifications. On the basis of these tests, the charcoal beds should perform ;

R. E. Adams; W. E. Browning

1958-01-01

38

The behavior of fission products during nuclear rocket reactor tests  

NASA Astrophysics Data System (ADS)

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

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

1991-01-01

39

Time dependent particle emission from fission products  

SciTech Connect

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

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

2010-01-01

40

Fission product release from core-concrete melts  

SciTech Connect

Based on measurements with a He-H2O carrier gas, release fraction values for lanthanum, barium, and strontium gaseous fission product species have been obtained from small-scale core-concrete melts at 2150 and 2385K. The sample configuration consisted of a layer of limestone or basaltic concrete granules, a layer of urania fuel granules containing La2O3, BaO, and SrO in solid solution, and a metal layer consisting of a mixture of stainless steel and Zircaloy between the fuel and concrete. The following trends in fission product release fraction values were found: Ba > Sr > La. Release fraction values based on measured release of fuel, cladding, structural, and concrete species are presented.

Tetenbaum, M.; Fink, J.K.; Johnson, C.E.; Ritzman, R.L.

1986-01-01

41

EVALUATION OF ACTIVATED CHARCOAL FISSION GAS ADSORBERS DESIGNED FOR THE GC ORR LOOP EXPERIMENT NO. 1  

Microsoft Academic Search

BS>The activated charcoal traps designed for the GasCooled Oak Ridge ; Research Reactor Loop Experiment No. 1 are evaluated for room temperature ; operation involving the decontamination and atmospheric disposal of helium ; coolant gas contaminated by experiment failure. The maximum quantity of fission ; products expected to be released was calculated and the resulting hazard was ; examined on

R. E. Adams; W. E. Browning

1960-01-01

42

Energy Dependence of Plutonium Fission-Product Yields  

NASA Astrophysics Data System (ADS)

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

Lestone, J. P.

2011-12-01

43

Application of the Mass Spectrometry Technique for Determination of the Gas Phase Composition in Irradiated Fuel Elements and Studying the Gas Fission Product Behaviour During Fuel Reprocessing.  

National Technical Information Service (NTIS)

The gas-phase composition is investigated in BOR-60 reactor irradiated fuel elements and in experimental fuel elements irradiated in the SM-2 reactor. The technique of mass-spectrometric analysis of the gas phase composition in irradiated fuel elements wi...

A. P. Kirillovich Y. I. Pimonov O. V. Skiba O. S. Boiko G. I. Vasil'ev

1980-01-01

44

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

45

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

SciTech Connect

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

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

2005-11-01

46

A model for nonvolatile fission product release during reactor accident conditions  

SciTech Connect

An analytical model has been developed to describe the release kinetics of nonvolatile fission products (e.g., molybdenum, cerium, ruthenium, and barium) from uranium dioxide fuel under severe reactor accident conditions. This treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of antimony and tellurium in the Zircaloy cladding. This model interprets the release behavior of fission products observed in Commissariat a l`Energie Atomique experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere.

Lewis, B.J.; Andre, B.; Ducros, G. [CEA Centre d`Etudes Nucleaires, Grenoble (France); Maro, D. [CEA Centre d`Etudes Nucleaires, Fontenay aux Roses (France)

1996-10-01

47

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS  

DOEpatents

The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

Nicholls, C.M.; Wells, I.; Spence, R.

1959-10-13

48

(Fission product transport processes in reactor accidents)  

SciTech Connect

The purpose of this trip was to participate in and to hold informal discussions with other participants in the International Centre for Heat and Mass Transfer (ICHMT) International Seminar on Fission Product Transport Processes held at Dubrovnik, Yugoslavia, during the week of May 22--26, 1989. There were 129 participants from 20 countries at the Seminar. The travelers delivered two invited lectures and presented four invited papers based upon NRC-sponsored work at Oak Ridge National Laboratory. One of the travelers also served as Chairman of the Session entitled Transport Phenomena in the Reactor Coolant System'' and appeared as a Panelist in the Closing Session of the Seminar.

Hodge, S.A.; Beahm, E.C.; Kress, T.S.; Malinauskas, A.P.

1989-06-14

49

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS  

DOEpatents

A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

1958-10-01

50

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

SciTech Connect

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

Acharya, R.; Hanson, D.

1988-05-01

51

Ceramic Hosts for Fission Products Immobilization  

SciTech Connect

Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent metal ions, Mg and Ca, in the ceramic host phases. The immobilization of rear earth (lanthanide series) fission products in these ceramic host phases will also be studied this year. Cerium oxide is chosen to represent the rear earth fission product for substitution studies in spinel, perovskite and zirconolite ceramic hosts. Cerium has +3 and +4 oxidation states and it can replace some of the trivalent or tetravalent host ions to produce the substitution ceramics such as MgAl2-xCexO4, CaTi1-xCexO3, CaZr1-xCexTi2O7 and CaZrTi2-xCexO7. X-ray diffraction analysis will be used to compare the crystalline structures of the pure ceramic hosts and the substitution phases. SEM-EDX analysis will be used to study the Ce distribution in the ceramic host phases. The range of cerium doping is planned to reach the full substitution of the trivalent or tetravalent ions, Al, Ti and Zr, in the ceramic host phases.

Peter C Kong

2010-07-01

52

Fission gas release in LWR fuel measured during nuclear operation  

SciTech Connect

A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods and to a gamma spectrometer. The measurements of Xe and Kr are made during nuclear operation at steady state power, and for /sup 135/I following reactor scram. The first experiments were conducted at a burnup of 3000 MWd/t UO/sub 2/, at bulk average fuel temperatures of approx. 850 K and approx. 23 kW/m rod power. The measured release-to-birth ratios (R/B) of Xe and Kr are of the same magnitude as those observed in small UO/sub 2/ specimen experiments, when normalized to the estimated fuel surface-to-volume ratio. Preliminary analysis indicates that the release-to-birth ratios can be calculated, using diffusion coefficients determined from small specimen data, to within a factor of approx. 2 for the IFA-430 fuel. The release rate of /sup 135/I is shown to be approximately equal to that of /sup 135/Xe.

Appelhans, A.D.; Skattum, E.; Osetek, D.J.

1980-01-01

53

Fission-product gamma-ray and photoneutron spectra  

Microsoft Academic Search

Fission-product gamma-ray and photoneutron spectra from thermal and fast fission of ²³³U, ²³⁵U, ²³⁸U, and ²³⁹Pu have been calculated at 27 time intervals between 1 and 1000 hours following reactor shutdown. The gamma spectral calculations were made using CINDER, a depletion and fission-product code, which has been revised, extended, and variably dimensioned for applications to many problems involving irradiated materials.

M. G. Stamatelatos; T. R. England

1975-01-01

54

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

Microsoft Academic Search

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

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

2010-01-01

55

Gamma-Ray Spectra of Fractionated Fission Products.  

National Technical Information Service (NTIS)

To determine the effects of fractionation on gamma-ray exposure rates in fission-product fields, spectra of gamma-rays emitted by fractionated products of thermal neutron fission of 235U were studied. Controlled fractionation was brought about by sweeping...

D. Sam L. R. Bunney

1971-01-01

56

SEPARATION OF FISSION PRODUCTS FROM PLUTONIUM BY PRECIPITATION  

DOEpatents

Fission product separation from hexavalent plutonium by bismuth phosphate precipitation of the fission products is described. The precipitation, according to this invention, is improved by coprecipitating ceric and zirconium phosphates (0.05 to 2.5 grams/liter) with the bismuth phosphate.

Seaborg, G.T.; Thompson, S.G.; Davidson, N.R.

1959-09-01

57

Concentration of Mixed Fission Products from Seawater by Chelex 100.  

National Technical Information Service (NTIS)

A study was conducted to determine the feasibility of concentrating mixed fission products in seawater by ion-exchange. Chelex 100, a chelating resin, was found to be effective for the concentration of several gamma-emitting fission-product radionuclides ...

M. G. Lai H. A. Goya

1967-01-01

58

Modeling fission product release from ruptured LWR fuel rods  

Microsoft Academic Search

The principal objectives of the fission product release program are to determine the quantity of radiologically significant fission products released from defected LWR fuel rods under accident conditions, identify their chemical and physical forms, and interpret the results for use as input to computer models of postulated transportation and loss-of-coolant accidents. Experimental work with flowing steam in the temperature range

R. A. Lorenz; J. L. Collins; A. P. Malinauskas

1978-01-01

59

Release of fission products from irradiated aluminide fuel at high temperature  

SciTech Connect

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

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

1982-01-01

60

(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

61

The growth of fission gas bubbles in irradiated uranium dioxide  

Microsoft Academic Search

The growth of fission gas bubbles from supersaturated solution in irradiated uranium dioxide has been studied by electron microscopy under isothermal annealing conditions between 1300 and 1500C. Measurements of the kinetics of bubble growth have enabled the diffusion coefficients of atomic xenon and krypton in irradiated uranium dioxide to be determined. The diffusion coefficients obtained may be expressed by the

R. M. Cornell

1969-01-01

62

Fission gas retention and axial expansion of irradiated metallic fuel  

SciTech Connect

Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin.

Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

1986-05-01

63

A computer model for predicting transient fission gas release from UO sub 2 fuel  

Microsoft Academic Search

In this paper, the KFGR-T computer model is developed to predict transient fission gas release from UO fuel with an emphasis on the nonequilibrium behavior of fission gas bubbles. It takes into account the relevant physical processes generally considered by other workers, as well as migration of fission gas bubbles through channels formed by the extension of dislocations to grain

K. S. Sim; H. C. Suk; Y. K. Yoon

1992-01-01

64

EVALUATION OF FISSION GAS ADSORPTION TRAPS FOR ORNL-MTR-44 LOOP EXPERIMENT  

Microsoft Academic Search

A method of predicting the performance of fission gas adsorption traps ; containing activated charcoal is presented. This method is applied in the ; evaluation of the fission gas traps designed for use in the ORNL-MTR44 loop ; experiment. The method should also be applicable in evaluating fission gas traps ; contained in other reactor experiments. (auth);

R. E. Adams; W. E. Browning

1958-01-01

65

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

NASA Astrophysics Data System (ADS)

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

Campbell, Joann Marie

1997-07-01

66

Search for production of superheavy elements via fusion after instantaneous fission in the reaction 238 U+ 208 Pb  

Microsoft Academic Search

Superheavy elements (SHE) might be formed via a reaction mechanism called fusion after instantaneous fission, which is supposed to occur during the collision of a deformed very heavy nucleus with a spherical one. We bombarded natural uranium targets with lead ions and searched for alpha-emitting and spontaneously fissioning reaction products. Different techniques were used: a rotating wheel, a gas jet

T. Lund; D. Hirdes; H. Jungclas; D. Molzahn; P. Vater; R. Brandt; P. Lemmertz; R. Fa; H. Wollnik; H. Gggeler

1981-01-01

67

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

SciTech Connect

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

Stansfield, O.M.

1990-03-15

68

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

Microsoft Academic Search

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF\\/B-VI evaluation by England and Rider, for the forthcoming ENDF\\/B-VII.1 database release.11We intend to release the ENDF\\/B-VII.1 database in December 2011, and all released data are subject to CSEWG approval. It is possible that the released evaluated data will differ from

M. B. Chadwick; T. Kawano; D. W. Barr; M. R. Mac Innes; A. C. Kahler; T. Graves; H. Selby; C. J. Burns; W. C. Inkret; A. L. Keksis; J. P. Lestone; A. J. Sierk; P. Talou

2010-01-01

69

Chemistry of fission product iodine under nuclear reactor accident conditions  

SciTech Connect

The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

Malinauskas, A.P.; Bell, J.T.

1986-01-01

70

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

71

Isotopic fission product release from nuclear fuel under severe core damage accident conditions  

SciTech Connect

Isotopic fission gas release behavior during SFD tests 1-1, 1-3, and 1-4 is strongly dependent on the pre-test fuel history. For SFD 1-1, where the majority of all the fission products were generated during the preconditioning period, very little difference in isotopic release behavior between short- and long-lived species is predicted. For the SFD 1-3 and 1-4 tests, where the majority of all short-lived fission gases decayed away during the 4-year cooling period, differences between the behavior of long- and short-lived species are predicted. Most of the intragranular fission product release has been shown to be due to a grain-growth/grain-boundary-sweeping mechanism. In addition, fuel liquefaction/dissolution processes can lead to increased release under these degraded-core-accident conditions. These predictions follow the trend of the observed phenomena.

Rest, J.; Osetek, D.J.; Hartwell, J.K.

1985-09-01

72

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

SciTech Connect

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

J M Harp; P D Demkowicz; S A Ploger

2012-10-01

73

Modelling of fission gas release and gaseous swelling of light water reactor fuels  

Microsoft Academic Search

Modelling to investigate the behavior of fission gas atoms of light water reactor fuels is elaborated. The model features the treatment of the grain boundary assumed to consist of two zones for solute gas atoms and intergranular bubbles. This, along with the athermal diffusion of fission gas atoms, enables the retention of gas during the base irradiation and the instantaneous

Toshiaki Kogai

1997-01-01

74

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

NASA Astrophysics Data System (ADS)

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

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

1994-07-01

75

ENERGY RELEASE FROM THE DECAY OF FISSION PRODUCTS  

Microsoft Academic Search

The beta-ray energy release, gamma-ray energy release, and gamma-ray ; energy spectrum were calculated as a function of time after shutdown for fission ; products due to thermal neutron fission of U²³⁵. Calculations were carried ; out for reactor operating times of 100 hours, 1000 hours, and for an infinite ; operating time. Only the results for infinite operating time

Speigler

1959-01-01

76

ENERGY RELEASE FROM THE DECAY OF FISSION PRODUCTS  

Microsoft Academic Search

The total disintegration rates, rates of beta- and gamma-energy release, ; and gamma-ray energy spectrum, are calculated for fission products due to thermai ; neutron fission of U²³⁵. Reactor operating times of 1, 10, 100, and 1000 ; hours are treated, and the results plotted for decay times ranging from 10² ; to 10⁸ seconds. In addition, results for instantaneous

J. F. Perkins; R. W. King

1958-01-01

77

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

Microsoft Academic Search

Behavior of helium and fission gases in irradiated mixed oxide (MOX) fuels was investigated by pin puncture and heating tests by quantitatively measuring amounts of helium and fission gases in a fuel pin irradiated in JOYO to ?50MWdkg?1 as a whole pin average burnup. While the fission gas releases were 47% and 48% for Kr and Xe respectively, all helium

I. Sato; K. Katsuyama; Y. Arai

2011-01-01

78

The behavior of fission products during nuclear rocket reactor tests  

NASA Astrophysics Data System (ADS)

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

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

79

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

80

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

81

Fission product removal from molten salt using zeolite  

SciTech Connect

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

Pereira, C.; Babcock, B.D.

1996-10-01

82

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

SciTech Connect

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

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

2004-09-30

83

Fission  

NASA Astrophysics Data System (ADS)

The breaking apart of a body into smaller fragments. In the context of nuclear physics the term, nuclear fission, refers to the splitting of a heavy atomic nucleus into two or more lighter nuclei with the release of energy. The mass of the nucleus prior to fission is greater than the combined masses of the fragments, the difference in mass, ?m, being released as a quantity of energy, ?E (?E=?mc2,...

Murdin, P.

2000-11-01

84

Deposition of volatile fission products during commercial high-level waste vitrification  

SciTech Connect

High-level waste from spent commercial PWR fuel was vitrified during the Nuclear Waste Vitrification Project (NWVP), conducted at Pacific Northwest Laboratory in 1978 and 1979. The solidification equipment employed sintered metal filters and a conventional wet off-gas cleanup system to decontaminate the gases generated during waste processing. This report is a summary of a study completed under contract with U.S. Department of Energy (DOE) to determine the fate of volatile fission products in the filters and off-gas system. The study objectives were to determine (1) the extent of volatile fission product deposition within the sintered metal filters, (2) the distribution of fission products in the off-gas system, and (3) the factors that control the distribution. Such information is needed for the design and operation of future waste immobilization systems. The scope of the study included identification, preparation, and microprobe analyses of suitable filter samples and radiochemical analyses of off-gas system samples. The report describes the methods of sample selection and gathering, the analytical methods and results, and the interpretations of the results. Relationships to vitrification system operations are defined and operating conditions to minimize problems recommended. 37 figures, 14 tables.

Hanson, M.S.; Carter, J.G.

1982-09-01

85

(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

86

Fission gas behavior in mixed-oxide fuel during transient overpower. [LMFBR  

Microsoft Academic Search

Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents\\/s) tested mixed-oxide fuel pins and their steady state siblings.

E. H. Randklev; H. A. Treibs; B. Mastel; D. L. Baldwin

1979-01-01

87

MARGARET: A comprehensive code for the description of fission gas behavior  

Microsoft Academic Search

The relevant phenomena concerning stable-fission gas behavior in nuclear fuels are combined in a single model: MARGARET. This same tool can be used for base irradiations up to high burnup, ramp tests and annealing tests. The representation of intragranular or intergranular bubbles and fabrication pores is highly mechanistic. The partition of fission gas between these cavities and dissolved in the

L. Noirot

2011-01-01

88

Acoustic Sensor for In-Pile Fuel Rod Fission Gas Release Measurement  

Microsoft Academic Search

Innovative in-pile instrumentation is crucial for ad- vanced experimental programs in research reactors. In this field, we developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in Material Testing Reactors (MTR). In order to perform experimental programs related to the study of the fission gas release kinetics,

D. Fourmentel; J. F. Villard; J. Y. Ferrandis; F. Augereau; E. Rosenkrantz; M. Dierckx

2011-01-01

89

Acoustic sensor for in-pile fuel rod fission gas release measurement  

Microsoft Academic Search

Innovative in-pile instrumentation is crucial for advanced experimental programs in research reactors. this field, we developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French

D. Fourmentel; J. F. Villard; J. Y. Ferrandis; F. Augereau; E. Rosenkrantz; M. Dierckx

2009-01-01

90

Characterization of intergranular fission gas bubbles in U-Mo fuel  

Microsoft Academic Search

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

Y. S. Kim; G. Hofman; J. Rest; G. V. Shevlyakov; SSCR RIAR

2008-01-01

91

Beta and Gamma Spectra of Short-Lived Fission Products  

Microsoft Academic Search

Continuous spectra of beta particles and gamma rays emitted in the decay of short-lived fission products have been measured. These spectra, which cover the complete energy range available, can be used for checking results of detailed spectroscopic work on the decay of the nuclides studied. Another application of basic nature is the use of the beta spectra to evaluate the

G. Rudstam; P. I. Johansson; O. Tengblad; P. Aagaard; J. Eriksen

1990-01-01

92

Gamma-ray spectrum of short lived fission products  

Microsoft Academic Search

The fission products from uranium-235 irradiated by slow neutrons have been analysed using a gamma-ray scintillation spectrometer. The spectra between 40 minutes and 40 hours after irradiation for 30 seconds have been recorded and twelve major activities identified.

J R Keane

1962-01-01

93

Decontamination of actinides and fission products from stainless steel surfaces  

Microsoft Academic Search

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,

C. Mertz; D. B. Chamberlain; L. Chen; C. Conner; G. F. Vandegrift; D. Drockelman; M. Kaminski; S. Landsberger; J. Stubbins

1996-01-01

94

Fission Product Activities in Aged Savannah River Plant Waste Solutions.  

National Technical Information Service (NTIS)

Fission products with activities equal to or greater than 10 nCi/ml were identified in 10-year-old defense waste solutions stored at the Savannah River Plant. Measured activities of exp 137 Cs, exp 106 Ru, exp 90 Sr, exp 99 Tc, and lanthanide isotopes pro...

J. R. Wiley

1978-01-01

95

LOW ENERGY NEUTRON CROSS SECTIONS OF RADIOACTIVE FISSION PRODUCT NUCLIDES  

Microsoft Academic Search

Data on Sm¹⁵¹ and Pm¹⁴⁷ were obtained, and resonances were ; assigned. Transmission measurements were also made on fission product samples of ; Zr and I, but no resonances could definitely be assigned to I¹²⁹ and Zr\\/sup ; 93\\/ in the energy range from 1 to 100 ev. (M.H.R.);

J. A. Harvey; R. C. Block; G. G. Slaughter; W. J. Martin; G. W. Parker

1958-01-01

96

Ab initio modelling of volatile fission products in uranium mononitride  

NASA Astrophysics Data System (ADS)

Defects and the incorporation of volatile fission products (xenon, krypton, caesium and iodine) in uranium mononitride are investigated using DFT calculations. Various locations for impurities are considered including at a tetrahedral interstitial position, substitution of a host nitrogen or uranium ion and placed in a Schottky defect (UN bivacancy). The incorporation is energetically more favourable for the latter, although the incorporation energies are positive. The preferred position for volatile fission products in UN is at the larger of the vacancies, either a single uranium vacancy or the uranium vacancy of a Schottky defect. The incorporation of a fission product in a bound [1 0 0]-Schottky defect leads to a tetragonal distortion of the supercell. The impurities considered in this work produce very small perturbations of the crystalline matrix of UN. With the exception of impurities at the interstitial site, which perturb the structure into the second coordination sphere, only the displacement of the atoms at the nearest-neighbour positions is significant. Analysis of the charge distribution after incorporation of the fission product reveals a weak charge transfer for the noble gases, while a larger transfer is displayed for caesium and iodine.

Klipfel, M.; van Uffelen, P.

2012-03-01

97

Average beta and gamma Energies of Fission Products.  

National Technical Information Service (NTIS)

The average beta and gamma energies in the decay of 382 known fission products have been determined. As far as possible the values are based on experimental data: direct determinations, published decay schemes, and a study of beta strength functions. In c...

G. Rudstam K. Aleklett

1979-01-01

98

Data summary report for fission product release test VI-5  

SciTech Connect

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

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

1991-10-01

99

Applications for fission product data to problems in stellar nucleosynthesis  

SciTech Connect

A general overview of the nucleosynthesis mechanisms for heavy (A greater than or equal to 70) nuclei is presented with particular emphasis on critical data needs. The current state of the art in nucleosynthesis models is described and areas in which fission product data may provide useful insight are proposed. 33 references, 10 figures.

Mathews, G.J.

1983-10-01

100

Benchmark experiments for fission product data (nuclear date)  

Microsoft Academic Search

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

Schenter

1975-01-01

101

Fission product scrubbing system for a nuclear reactor  

Microsoft Academic Search

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

1986-01-01

102

Simultaneous Estimation of Pu and Fission Products by Gamma Spectrometry.  

National Technical Information Service (NTIS)

A gamma spectrometric method is described for the simultaneous estimation of Pu and fission products using a 62cc intrinsic germanium detector coupled to a 4K MCA. The 120 KeV peak of /sup 239/Pu was employed for the assay of plutoniu m. The low energy 51...

P. V. Achuthan R. G. Bhogale A. Ramanujam M. R. Iyer D. N. Sharma

1988-01-01

103

Testing of Nuclear Data Libraries for Fission Products  

SciTech Connect

The status of the radiative capture and neutron inelastic scattering evaluations for the most important fission products is analyzed. New BROND-3 evaluations are briefly considered. Experiments on the BFS critical assemblies, which can be used to test the available evaluations, are discussed.

Ignatyuk, A.V.; Bednyakov, S.M.; Koshcheev, V.N.; Manokhin, V.N.; Manturov, G.N.; Tertuchny, G.Ya. [Institute of Physics and Power Engineering, 240020 Obninsk (Russian Federation)

2005-05-24

104

Testing of Nuclear Data Libraries for Fission Products  

NASA Astrophysics Data System (ADS)

The status of the radiative capture and neutron inelastic scattering evaluations for the most important fission products is analyzed. New BROND-3 evaluations are briefly considered. Experiments on the BFS critical assemblies, which can be used to test the available evaluations, are discussed.

Ignatyuk, A. V.; Bednyakov, S. M.; Koshcheev, V. N.; Manokhin, V. N.; Manturov, G. N.; Tertuchny, G. Ya.

2005-05-01

105

Calculation of fission products release from pebble-bed reactors  

Microsoft Academic Search

Thesis. Using a two-region diffusion model, fission product release ; from coated particles is calculated under simplified assumptions. With the help ; of Laplace-transforms the time-dependent diffusion equations are solved with ; suitable boundary conditions. A second solution is based on the application of ; Green's functions. Both methods yield the same results. Knowing the temperature ; distribution throughout the

Herhadi

1972-01-01

106

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

107

ORNL studies of fission product release under LWR accident conditions  

Microsoft Academic Search

High burnup Zircaloy-clad UO 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

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

1991-01-01

108

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

109

Fission product release as a function of chemistry and fuel morphology.  

National Technical Information Service (NTIS)

Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are ...

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

1989-01-01

110

Thermal release of volatile fission products from irradiated nuclear fuel  

SciTech Connect

An effective procedure for removing /sup 3/H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO/sub 2/ fuel. The release characteristics of /sup 3/H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released about 99% of the /sup 3/H and 20% of both Kr and Xe within a 3 h at 1500/sup 0/C. Similar results were obtained using a carrier gas of He containing 6% H/sub 2/. However, a carrier gas containing only He resulted in the release of approximately 80% of the /sup 3/H and 99% of both Kr and Xe. These results indicate that the release of these volatile fission products from irradiated nuclear fuel is a function of the chemical composition of the gaseous environment. The rate of tritium release increased with increasing temperature (1100 to 1500/sup 0/C) and with the addition of hydrogen to the gas stream. Using crushed UO/sub 2/ fuel without cladding and He as the carrier gas, Kr was completely released at 1500/sup 0/C in 2.5 h. Below 1350/sup 0/C, no Kr-Xe release was observed. Approximately 86% of the /sup 129/I and 95% of the cesium was released from a piece (3.9 g) of UO/sub 2/ fuel at 1500/sup 0/C in He. The zirconium cladding was observed to fracture during heat treatment. A large-scale thermal outgassing system was conceptually designed by the General Atomic Company from an engineering analysis of available experimental data. The direct cost of a 0.5 metric/ton day thermal outgassing system is estimated to be $1,926,000 (1982 dollars), including equipment, installation, instrumentation and controls, piping, and services. The thermal outgassing process was determined to be a technically feasible and cost-competitive process to remove tritium in the head-end portion of a LWR fuel reprocessing plant. Additional laboratory-scale development has been recommended.

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

1983-06-01

111

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

NASA Astrophysics Data System (ADS)

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release.1We intend to release the ENDF/B-VII.1 database in December 2011, and all released data are subject to CSEWG approval. It is possible that the released evaluated data will differ from those presented in this paper; the evaluated date presented here can be referred to as ENDF/B-VII.1 beta 0. We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small especially for 99Mo we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for 95Zr, 140Ba, 144Ce), but are larger for 99Mo (4%-relative) and 147Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the 147Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the energy dependence over the fast neutron energy range from 0.2-2 MeV. Based on these trends, we present an evaluation of the FPY data at 0.5 and 2.0 MeV average incident neutron energies. This new set of ENDF/B-VII data will enable users to linearly interpolate between the pooled FPY data at 0.5 MeV and our new data at 2 MeV to obtain FPYs at other energies.

Chadwick, M. B.; Kawano, T.; Barr, D. W.; Mac Innes, M. R.; Kahler, A. C.; Graves, T.; Selby, H.; Burns, C. J.; Inkret, W. C.; Keksis, A. L.; Lestone, J. P.; Sierk, A. J.; Talou, P.

2010-12-01

112

Fission product iodine and cesium release behavior under light water reactor accident conditions  

Microsoft Academic Search

The release behavior of the fission products iodine and cesium has been characterized in fission product release tests that have been conducted at Oak Ridge National Laboratory with highly irradiated, light water reactor fuel segments under conditions simulating severe accidents. The chemical forms of the fission products depended on the composition of the carrier gases used in the tests. In

J. L. Collins; M. F. Osborne; R. A. Lorenz; A. P. Malinauskas

1988-01-01

113

Identification of {mu}s isomers in fission products  

SciTech Connect

Several {mu}s isomers have been observed in the fission products of {sup 241}Pu and {sup 239}Pu. The detection is based on the time correlation between reaction products selected by the LOHENGRIN spectrometer and the {gamma}-rays depopulating the isomers. New isomers have been observed in {sup 96}Rb, {sup 106}Nb, {sup 127}Sn, and {sup 130}Te and the others have been confirmed in {sup 94}Y, {sup 126}Sn, {sup 129}Sn and {sup 129}Sb.

Genevey, J.; Pinston, J. A.; Ibrahim, F.; Larque, T. [Institut des Sciences Nucleaires, 38026 Grenoble (France); Faust, H.; Friedrichs, T.; Gross, M.; Oberstedt, S. [Institut Laue Langevin, 38042 Grenoble (France)

1998-12-21

114

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

SciTech Connect

The mass and charge distributions in an unseparated mix of fission product nuclei from thermal-neutron fission of /sup 242m/Am were studied through semiconductor gamma-ray spectrometry. Samples of the fissionable material under study were irradiated in a vertical irradiation tube of the MIFI IRT research reactor. Following irradiation, measurements were made on aperture-calibrated semiconductor detectors. For broader identification of fission fragment nuclides three experiments were conducted that differed substantially in irradiation duration. The spectrum of gamma radiation from the mix of fission products and the time dependences of count rate at total absorption peaks were analyzed on SM-4 and Iskra-226 computers. The values of yields obtained were compared with data of investigations conducted earlier with other experimental methods, and also with the results of calculations.

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

1985-03-01

115

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

Microsoft Academic Search

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

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

1990-01-01

116

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

Microsoft Academic Search

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

R. A. Anderl; R. C. Greenwood

1990-01-01

117

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

SciTech Connect

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

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

1987-01-01

118

Transmutation of selected fission products in a fast reactor  

SciTech Connect

A small fast reactor such as the Fast Flux Test Facility can be an effective device for destroying long-lived fission products such as {sup 99}Tc and {sup 129}I. There are several potential core configuration options using both fast and moderated neutron spectra. The calculated Doppler reactivity coefficient for {sup 99}Tc was 60% of the value for {sup 238}U on a per atom basis. Replacing {sup 238}U with {sup 99}Tc in waste burn applications has the positive attributes of reduced parasitic capture in uranium and enhanced fission product destruction, while retaining a substantial Doppler effect. A modular liquid metal reactor system could support about 8 to 10 comparably sized conventional light water reactors.

Wootan, D.W.; Nelson, J.V.

1993-05-01

119

Re-solution of fission gas A review: Part I. Intragranular bubbles  

NASA Astrophysics Data System (ADS)

Theories of fission-fragment-driven re-solution of fission-gas atoms from intragranular bubbles in irradiated UO2 nuclear fuel are reviewed. Two mechanisms of re-solution are generally accepted: the heterogeneous process destroys entire bubbles in the path of fission fragments and returns the gas to the solid as individual atoms; the homogeneous process re-solves fission-gas atoms singly by scattering collisions with fission fragments and uranium recoils whose paths intersect the bubbles. Coupling of these two re-solution models with the bubble nucleation analogs determines the size and number density of the intragranular bubble population. Two approaches are reviewed: the single-size theory, in which all bubbles are accorded one size, and the bubble distribution theory, which seeks to determine the variation of bubble number density with size.

Olander, D. R.; Wongsawaeng, D.

2006-08-01

120

[Chronic radiation damage by uranium and plutonium nuclear fission products].  

PubMed

Chronical radiation sickness is a special form of the radiation damage. It occurs when doses of chronical irradiation exceed their values established for professionals. The sickness is well studied in clinical observations. It may take place also if uranium and plutonium nuclear fission products (NFP) enter the organism. In the last case the chronical radiation sickness practically is not investigated. In the article we present the results of the experimental studies on dogs of the damage caused by NFP. PMID:17323703

Vasilenko, I Ia; Vasilenko, O I

121

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

122

Fission product release from fuel under severe accident conditions  

Microsoft Academic Search

Recent advances in the understanding of fission product release from fuel under severe accident conditions in light water reactors are reviewed. In addition to the effects of temperature and time at temperature, recent results from in-pile and out-of-pile tests and the accident at Three Mile Island Unit 2 suggest that the effects of fuel morphology such as restructuring of the

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

1993-01-01

123

Fission product source terms and engineered safety features  

SciTech Connect

The author states that new, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents will soon be available. Although these methodologies will undoubtedly find widespread use in the development of accident response procedures, the author states that it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to strategies for the mitigation of fission product releases. Questions concerning the performance of existing engineered safety systems are reviewed.

Malinauskas, A.P.

1984-01-01

124

Review of Neutron Cross-Section Evaluations for Fission Products  

SciTech Connect

Review of neutron cross-section evaluations for fission products included in 5 major evaluated nuclear-data libraries was performed. The aim of the project, conducted under the WPEC Subgroup 21 during 2001-2004, was to prepare recommendations for best evaluations of nuclei in the range Z 31-68. Altogether, existing evaluations for 211 materials were reviewed, 7 new materials were added, and recommendations were prepared for 218 materials.

Oblozinsky, P.; Herman, M.; Mughabghab, S. [National Nuclear Data Center, BNL, Upton, NY 11973 (United States); Sirakov, I. [National Nuclear Data Center, BNL, Upton, NY 11973 (United States); INRNE, Sofia (Bulgaria); Chang, J. [Nuclear Data Laboratory, KAERI, Tajeon (Korea, Republic of); Nakagawa, T.; Shibata, K. [Nuclear Data Center, JAERI, Tokai-mura (Japan); Kawai, M. [KEK, Tsukuba (Japan); Ignatyuk, A.V. [Laboratory of Theoretical Physics, IPPE, Obninsk (Russian Federation); Pronyaev, V.G.; Zerkin, V. [Nuclear Data Section, IAEA, Vienna (Austria); Shen Qingbiao; Zhuang Youxiang [China Nuclear Data Center, Beijing (China)

2005-05-24

125

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.

126

Measurements for the JASPER Program fission gas plenum experiment  

SciTech Connect

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

Muckenthaler, F.J.

1987-06-01

127

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

128

PHASE I REPORT OF DEVELOPMENT TECHNIQUES FOR POWER PRODUCTION FROM MIXED FISSION PRODUCTS  

Microsoft Academic Search

An investigation was made into the various processes for the fixation of ;\\u000a mixed fission products as solids in order to determine the extent they could be ;\\u000a utilized as heat sources for thermoelectric generators. Generators of up to ten ;\\u000a watts can be designed and built with state-of-art'' thermoelectric materials ;\\u000a and mixed fission products soon to be available

1961-01-01

129

Highlights of the OECD LOFT LP-FP-2 experiment including hydrogen generation, fission product chemistry, and transient fission product release fractions.  

National Technical Information Service (NTIS)

The purpose of this paper is to present the primary highlights of the OECD LOFT Experiment LP-FP-2. These highlights include the chemistry of the released fission products, the transient fission product release fractions, and the total amount and distribu...

M. L. Carboneau

1990-01-01

130

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

Microsoft Academic Search

A Monte Carlo method for the estimation of ?-delayed ?-ray spectra following fission is described that can accomodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each

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

2004-01-01

131

Grain Boundary Percolation Modeling of Fission Gas Release in Oxide Fuels  

SciTech Connect

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

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

2012-05-01

132

Fission-Product Gamma-Ray and Photoneutron Spectra and Energy-Integrated Sources Informal Report.  

National Technical Information Service (NTIS)

Fission-product gamma-ray and photoneutron spectra resulting from thermal and/or fast fission of 232Th, 233U, 235U, 238U, 239Pu, and 241Pu have been calculated at a number of times following fission. Reported here are spectra at 1, 10, 100, 1000, and 5000...

M. G. Stamatelatos T. R. England

1977-01-01

133

The Effects of the Depletion and Buildup of Fissile Nuclides and of 238U Fast Fissions on Fission Product Decay Power.  

National Technical Information Service (NTIS)

The major source of power in the first 0 to 10,000 seconds after shutdown in a light water reactor (LWR) is the decay power of the fission products. The total decay power in LWR's result from fission products formed in thermal fissions of 235U, 239Pu, and...

T. J. Trapp

1978-01-01

134

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

135

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

136

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

NASA Astrophysics Data System (ADS)

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

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

2005-05-01

137

PURIFICATION OF PLUTONIUM USING A CERIUM PRECIPITATE AS A CARRIER FOR FISSION PRODUCTS  

DOEpatents

Bismuth phosphate carrier precipitation processes are described for the separation of plutonium from fission products wherein in at least one step bismuth phosphate is precipitated in the presence of hexavalent plutonium thereby carrying a portion of the fission products from soluble plu tonium values. In this step, a cerium phosphate precipitate is formed in conjunction with the bismuth phosphate precipitate, thereby increasing the amount of fission products removed from solution.

Faris, B.F.; Olson, C.M.

1961-07-01

138

ORNL studies of fission product release under LWR severe accident conditions  

Microsoft Academic Search

The large inventories of radioactive fission products in irradiated fuel represent the principal personnel hazards from nuclear reactors. A large fraction of the existing fission-product release data has been collected from experiments at Oak Ridge National Laboratory (ORNL). Tests of high-burnup light-water-reactor fuel, and also simulated fuel with fission-product tracers, have been conducted in an induction furnace at temperatures up

M. F. Osborne; R. A. Lorenz

2009-01-01

139

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

140

Permian basin gas production  

SciTech Connect

Of the 242 major gas fields in the Permian basin, 67 are on the Central Basin Platform, 59 are in the Delaware basin, 44 are in the Midland basin, 28 are in the Val Verde basin, 24 are on the Eastern Shelf, 12 are in the Horshoe Atoll and eight are on the Northwest Shelf. Eleven fields have produced over one trillion cubic feet of gas, 61 have produced between 100 billion and one trillion cubic feet of gas and 170 have produced less than 100 billion cubic feet. Highlights of the study show 11% of the gas comes from reservoirs with temperatures over 300 degrees F. and 11% comes from depths between 19,000 and 20,000 feet. Twenty percent of the gas comes from reservoirs with pressures between 1000 and 2000 psi, 22% comes from reservoirs with 20-24% water saturation and 24% comes from reservoirs between 125 and 150 feet thick. Fifty-three reservoirs in the Ellenburger formation have produced 30% of the gas, 33% comes from 88 reservoirs in the Delaware basin and 33% comes from reservoirs with porosities of less than five percent. Forty percent is solution gas and 46% comes from combination traps. Over 50% of the production comes from reservoirs with five millidarcys or less permeability, and 60% of the gas comes from reservoirs in which dolomite is the dominant lithology. Over 50% of the gas production comes from fields discovered before 1957 although 50% of the producing fields were not discovered until 1958.

Haeberle, F.R. [Consulting Geologist, Dallas, TX (United States)

1995-06-01

141

Studies of fission product movement in tuffaceous media  

SciTech Connect

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

Thompson, J.L.

1991-09-01

142

Heat and fission product transport in molten core material pool with crust  

Microsoft Academic Search

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products

J. I. Yun; K. Y. Suh; C. S. Kang

2005-01-01

143

Search for production of superheavy elements via ``fusion after instantaneous fission'' in the reaction238U+208Pb  

NASA Astrophysics Data System (ADS)

Superheavy elements (SHE) might be formed via a reaction mechanism called fusion after instantaneous fission, which is supposed to occur during the collision of a deformed very heavy nucleus with a spherical one. We bombarded natural uranium targets with lead ions and searched for alpha-emitting and spontaneously fissioning reaction products. Different techniques were used: a rotating wheel, a gas jet system and a radiochemical procedure. No SHE have been found. The upper cross section limits are between 10-33 cm2 and 10-34 cm2 within the half-life range from 1 ms to 1 year. In addition, the production cross sections of some Cm and Cf isotopes and of Am fission isomers were determined.

Lund, T.; Hirdes, D.; Jungclas, H.; Molzahn, D.; Vater, P.; Brandt, R.; Lemmertz, P.; Fa, R.; Wollnik, H.; Gggeler, H.

1981-06-01

144

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

Microsoft Academic Search

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

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

1985-01-01

145

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

NASA Astrophysics Data System (ADS)

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

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

2011-09-01

146

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

147

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

SciTech Connect

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

Skim, Y. S.

1999-11-12

148

Identification of fissile materials from fission product gamma-ray spectra  

Microsoft Academic Search

A new activepassive nondestructive assay technique has been developed to identify fissile materials by observing the gamma-ray emissions from induced fission products. This technique entails using a neutron source to induce fissions in unknown containerized fissile materials and to subsequently observe high energy peaks (>800keV) in the fission product gamma-ray spectrum. Ratios of the observed peak intensities are then formed.

D. H Beddingfield; F. E Cecil

1998-01-01

149

JRC's on-line fission gas release monitoring system in the high flux reactor Petten  

Microsoft Academic Search

For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed, dubbed the Sweep Loop Facility (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experiments. The SLF

M. Laurie; M. A. Ftterer; K. H. Appelman; J. M. Lapetite; A. Marmier; G. Berg

150

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

151

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

2006-09-01

152

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

153

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

154

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

155

THE FISSION GAS PROBLEM FOR MOBILE FUEL FAST REACTORS  

Microsoft Academic Search

The problems that might result from the release of fission gases in ; mobile fuel fast reactors are considered for two types of mobile fuel systems; ; namely, a molten alloy fuel system of the type to be used in the Los Alamos ; Molten Plutonium Reactor Experiment and a paste fuel system of the type being ; developad by

F. G. Hammitt; E. C. Kovacic; F. J. Leitz

1960-01-01

156

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

Microsoft Academic Search

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

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

1988-01-01

157

Fission product release as a function of chemistry and fuel morphology  

Microsoft Academic Search

Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are the chemistry within the core and the morphology of the fuel or fuel-bearing debris. This paper reviews the current understanding of the these

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

1989-01-01

158

Utilization of fast reactor excess neutrons for burning long lived fission products  

Microsoft Academic Search

An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction.The fission product target assemblies which consist of

K Kawashima; K Kobayashi; K Kaneto

1995-01-01

159

Multigroup beta and gamma Spectra of Individual Endf/B-IV Fission-Product Nuclides.  

National Technical Information Service (NTIS)

Delayed beta and gamma group-energy spectra were calculated for 180 individual fission-product nuclides having spectral data in ENDF/B-IV fission-product files. The beta spectra, in uniform-grid 75-group structure, and the gamma spectra, in uniform-grid 1...

T. R. England M. G. Stamatelatos

1976-01-01

160

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

Microsoft Academic Search

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

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

161

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

Microsoft Academic Search

This report presents a study of fission product transport behavior in N Reactor during a severe accident. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The current report is an extension to a previous interum study that has added an aerosol formation model, replaced an older

Hagrman

1989-01-01

162

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

SciTech Connect

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

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

2010-12-15

163

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

164

Use of fission product implantation in nuclear waste management  

NASA Astrophysics Data System (ADS)

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

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

1998-10-01

165

SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES  

DOEpatents

Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

Maddock, A.G.; Booth, A.H.

1960-09-13

166

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

SciTech Connect

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

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

1999-05-19

167

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

168

Oil and gas production  

Microsoft Academic Search

Colloid and interface science is fundamental to many aspects of oil and gas production processes. This review focuses on recent advances in water-based drilling fluids, reservoir fracturing fluids, polymer gels for modifying reservoir permeability, avoiding problems caused by colloidal asphaltene oil fractions and characterising the wettability of reservoir rocks. There have been major advances in the design and responsiveness of

G. C Maitland

2000-01-01

169

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

Microsoft Academic Search

A technique has been developed to distinguish between ^{239}Pu and ^{235}U by observing fission product delayed gamma-rays produced by fissions induced by an external neutron source. If the number of induced fissions per source neutron per unit mass can be determined from Monte Carlo simulation, the material can also be quantified. Trials were performed with yellowcake, HEU-metal, and Pu-metal samples

David Harris Beddingfield

1996-01-01

170

Methods and results of an investigation of the distribution of solid fission products in HTGR fuel elements  

Microsoft Academic Search

A basic index of the reliability of HTGR fuel elements is their ability to retain fission products, by preventing them from penetrating into the primary circulation loop of the helium coolant. Here the limiting factor should be the yield of some of the solid fission products, because their half-lives are larger than those of gaseous fission products, and their activity

A. I. Deryugin; A. S. Chernikov; B. P. Kolesnikov; K. N. Koshcheev; A. D. Kurepin

1992-01-01

171

ESTIMATE OF HAZARD PRODUCED BY ACCIDENTAL RELEASE OF GASEOUS FISSION PRODUCTS FROM AN ORR FUSED SALT CAPSULE EXPERIMENT  

Microsoft Academic Search

An accidental release of gaseous fission products from an ORR fused salt ; capsule, containing 26 mg. of U²³⁵, was postulated and the resuiting hazard ; estimated by calculating the maximum external and internal dose an individual ; could receive from exposure to the gaseous fission products and their decay ; products. Assuming all the contained gaseous fission produets are

R. E. Adams; W. E. Browning

1959-01-01

172

Observation of isomer production in abrasion fission of ^238U on a ^9Be Target  

NASA Astrophysics Data System (ADS)

This talk will present the observation of gamma decay from isomeric states produced in abrasion fission of ^238U on a ^9Be target at 80 MeV/nucleon. This experiment was performed at the National Superconducting Cyclotron Laboratory at Michigan State University. The fission products were identified by A, Z and Q, with the gamma decay observed within a 20 microsecond window following implantation in a silicon telescope. This technique, for identification of breakup products is known as isomer tagging,. Isomer tagging has become an important tool for in-flight fragment identification of fission and fragmentation products. Unfortunately an extensive database of isomers is unavailable for much of the neutron rich region populated by fission. Because of this, one of the goals of fission studies at the NSCL has been to measure the population of isomeric states. These results along with the possible identification of previously unknown isomeric states will be presented.

Nettleton, A. S.; Amthor, A. M.; Folden, C. M., III; Ginter, T. N.; Hausmann, M.; Morrissey, D. J.; Portillo, M.; Sherrill, B. M.; Tarasov, O. B.; Kubo, T.; Nakao, T.; Takeda, H.; Loveland, W. D.; Manikonda, S. L.; Souliotis, G. A.

2007-10-01

173

Determining isotopic distributions of fission products with a Penning trap  

NASA Astrophysics Data System (ADS)

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

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

2010-04-01

174

Model predictions of the fission-product yields for 238 U Part IV: Selected case studies  

Microsoft Academic Search

The results of model calculations on nuclide yields produced in se- lected scenarios using reactions with 238 U nuclei are presented. The calculation combines the modelling of the initial reaction mechanism, the deexcitation proc- ess including the fission competition with the prediction of the nuclide production in the fission of excited 238 U and all daughter nuclei produced in the

K.-H. Schmidt; J. Benlliure; A. Junghans; B. Jurado; K. Helariutta; V. Ricciardi; J. Pereira; J. Taieb

2001-01-01

175

Analysis of the magnetohydrodynamic flow of a fissioning gas in a disk MHD generator  

SciTech Connect

The influence of fissioning and magnetohydrodynamic (MHD) interaction on the steady, supersonic flow of a compressible, turbulent, weakly ionized, fissioning gas in an outflow disk MHD generator is investigated in this work. The two-dimensional (r,z) MHD flow is modeled using the thin-layer Navier-Stokes equations with MHD and fission power density source terms, and Maxwell's equations under the MHD Approximations and assuming negligible induced magnetic induction. The simple plasma physics models used in this work suggest that the electron number densities (O 10[sup 19]m[sup 3]) and corresponding electrical conductivity levels (O 1 S/m) obtained from fission-fragment induced ionization alone may be insufficient for practical MHD generator operation. The MHD flow equations with the fission power density source term are integrated in boundary-fitted coordinates using the explicit method of MacCormack. The equations of electromagnetics, with variable plasma physics transport properties, are solved using an Alternating-Direction-Implicit (ADI) scheme. A consistent 2-D MHD solution is obtained by iteration between the fluid solver an the electromagnetics solver. The 2-D M solution methodology is used to analyze the influence of duct geometry and fission power density (for neutron flux levels between 0 and 10[sup 17] n/cm[sup 2]s) on the behavior of internal supersonic flows (with total Mach numbers less than 3), and to characterize the effects of variable applied magnetic induction levels and generators load resistances on the spatial profiles of important generator variables. THe predictions of the 2-D MHD solver developed in this work are compared with those of a quasi-one-dimensional Euler solver with MHD an fission source terms; the agreement between the two approaches suggests that the quasi-one-dimensional Euler solver does an excellent job predicting the behavior of supersonic, fissioning, disk MHD flows.

Welch, G.E.

1992-01-01

176

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

Microsoft Academic Search

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

Y. Zerega; C. Reynard-Carette; D. Parrat; M. Carette; B. Brkic; A. Lyoussi; G. Bignan; A Janulyte; J. Andre; Y. Pontillon; G. Ducros; S. Taylor

2011-01-01

177

Multiple voltage electron probe microanalysis of fission gas bubbles in irradiated nuclear fuel  

Microsoft Academic Search

The accurate analysis of locally retained fission gas in nuclear fuel is inherently difficult since the physical form under which it is stored varies from an atomic dispersion to bubbles with a diameter of several hundreds of nanometers. One of the techniques that has been applied since more than 20 yr is electron probe microanalysis (EPMA). This technique, however, is

M. Verwerft

2000-01-01

178

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

SciTech Connect

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

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

2012-02-01

179

Characterization of Individual Fission Products in Terms of Their Production and Transmutation  

SciTech Connect

In order to establish a simple and common basis which can be referred for the specific studies on the fission products (FP) transmutation and its strategies, the general characteristics of the dominant individual FPs in terms of their production and transmutation in the fast reactor have been studied with an ideal model in the present paper. The potential hazard of each nuclide in long-term utilization of nuclear energy in human society has been quantitatively evaluated. During utilization of fission energy, two short half-life FPs, {sup 90}Sr and {sup 137}Cs, almost determine the total toxicity of FP nuclides in spite of the effort of transmuting them in fast reactors. The innovative transmuters (such as High Flux Reactor, Accelerator-driven system and Fusion Neutron Source) are needed to reduce these 2 toxicities. In case of 1000-year fission energy utilization with transmutation in fast reactors, the total toxicity of FP nuclides except {sup 126}Sn go down below the level of {sup 238}U toxicity consumed. The more detailed study of transmutation of {sup 126}Sn is a very important in future. However, in the case of 10,000-year fission energy utilization with transmutation in fast reactor, total toxicities of all FP nuclides go down below the level of {sup 238}U toxicity consumed. (authors)

Hiroshi Sagara; Tadashi Yoshida; Masaki Saito [Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152 (Japan)

2002-07-01

180

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

SciTech Connect

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

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

1980-10-01

181

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

182

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

183

IONIZATION-RATE AND PHOTON PULSE-RATE DECAY OF FISSION PRODUCTS FROM THE SLOW-NEUTRON FISSION OF U²³⁵  

Microsoft Academic Search

Computations giving the theoretical ionization rate and response of the ; AN\\/PDR-39(TID) radiac at 3 feet above an infinite plane uniformly contaminated ; with fission products from 10⁴ fissions of U²³⁵ per sq ft. are ; presented as a function of time after fission. The response of the USNRDL 4 pi ; ionization chamber and the photon pulse rate for

C. F. Miller; P. Loeb

1958-01-01

184

Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors  

Microsoft Academic Search

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

Dawn Scates

2010-01-01

185

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

SciTech Connect

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

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

2008-12-15

186

Determining isotopic distributions of fission products with a Penning trap  

Microsoft Academic Search

A novel method to determine independent yields in particle-induced fission employing the ion guide technique and ion counting\\u000a after a Penning trap has been developed. The method takes advantage of the fact that a Penning trap can be used as a precision\\u000a mass filter, which allows an unambiguous identification of the fission fragments. The method was tested with 25MeV and

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

2010-01-01

187

Recrystallization and fission-gas-bubble swelling of U-Mo fuel  

NASA Astrophysics Data System (ADS)

At high burnup, U-Mo fuel exhibits some form of recrystallization, by which fuel grains are subdivided. The effect of grain subdivision is to effectively enhance fission gas bubble (FGB) swelling due to increased grain boundaries. Inter-granular FGB swelling, i.e., FGB formation and growth at the grain boundaries, is much larger than the intra-granular FGB swelling. Recrystallized fuel volume fractions of U-Mo fuels irradiated to fission densities reaching 5.7 1021 f/cm3 were measured. Analytical expressions of recrystallization kinetics of U-Mo fuel during irradiation have been developed through the usage of the Avrami equation, a phenomenological equation which is also used to describe similar typical transformation reactions, such as new phase formation. In this work, we present a novel FGB swelling model of U-Mo fuel that is expressed in terms of Mo content, extent of cold work (fuel powder fabrication method), and fission density.

Kim, Yeon Soo; Hofman, G. L.; Cheon, J. S.

2013-05-01

188

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

NASA Astrophysics Data System (ADS)

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

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

2013-03-01

189

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

National Technical Information Service (NTIS)

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

M. Bonnevie-Svendsen V. Martini

1966-01-01

190

Sorption of Radionuclides on Geologic Media - A Literature Survey. I: Fission Products.  

National Technical Information Service (NTIS)

The fission products investigated were cobalt, nickel, strontium, cesium, technetium and iodine. Parameters of importance to sorption have been identified and a tabulation of distribution coefficients for groundwater conditions (pH 7-9, low to medium ioni...

K. Andersson B. Allard

1983-01-01

191

Fission and Corrosion Products Behavior in Primary Circuits of LMFBR'S. Proceedings.  

National Technical Information Service (NTIS)

Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedur...

A. W. Thorley H. Feuerstein

1987-01-01

192

Evaluations of Fission Product Capture Cross Sections for ENDF/B-V.  

National Technical Information Service (NTIS)

Capture cross section evaluations were made for the 36 most important fission product absorbers in a fast reactor system. These evaluations were obtained by use of a generalized least-squares approach with calculations being performed with the computer co...

R. E. Schenter D. L. Johnson F. M. Mann F. Schmittroth

1979-01-01

193

ARSENATE CARRIER PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM NEUTRON IRRADIATED URANIUM AND RADIOACTIVE FISSION PRODUCTS  

DOEpatents

A process is described for precipitating Pu from an aqueous solution as the arsenate, either per se or on a bismuth arsenate carrier, whereby a separation from uranium and fission products, if present in solution, is accomplished.

Thompson, S.G.; Miller, D.R.; James, R.A.

1961-06-20

194

EMERGENCY MONITORING METHODS FOR THE DETERMINATION OF THE EFFECTIVE BONE SEEKING FISSION PRODUCTS IN MILK  

Microsoft Academic Search

A radiochemical-radiometric prccedure is described for determining bone-; seeking fission products in milk. The prccedures permit completion of a few ; samples by a single technician in 6 to 8 hours. (C.J.G.);

N. I. Sax; R. J. Sherer; R. T. Drew

1959-01-01

195

Delayed-neutron branching ratios of fission products. A status report.  

National Technical Information Service (NTIS)

Delayed-neutron branching ratios have been reviewed for 86 nuclides, including a few isomers, among the fission products. The list comprises values reported before the end of December, 1987. (authors) (33 refs.). (Atomindex citation 21:070031)

E. Lund G. Rudstam

1989-01-01

196

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

197

Nondestructive Measurement of the Radial Two-Dimensional Distributions of Fission Products in Irradiated Fuel Materials.  

National Technical Information Service (NTIS)

The radial two-dimensional isotopic distributions of fission products are determined by measuring diametral gamma scans or projections at two or more angular orientations and analytically unfolding these projections. This nondestructive examination techni...

J. R. Phillips B. K. Barnes M. L. Barnes

1982-01-01

198

Reconstruction of Radial Fission-Product Distributions in Reactor Fuels from a Small Number of Projections.  

National Technical Information Service (NTIS)

Four mathematical techniques for reconstruction of the radial two-dimensional distribution of fission products using projections obtained by nondestructive gamma scanning were evaluated. Reconstruction of a picture from a finite set of projections is math...

B. K. Barnes J. R. Phillips M. L. Barnes

1981-01-01

199

Fission Product Removal in Engineered Safety Feature (ESF) Systems: Data Base Assessment and Suggested Experimental Program.  

National Technical Information Service (NTIS)

The available data base on the fission product removal capabilities of nuclear reactor Engineered Safety Feature (ESF) systems was reviewed and assessed. The systems considered included pressure suppression pools, ice condenser systems, containment sprays...

F. R. Zaloudek A. K. Postma W. K. Winegardner

1984-01-01

200

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

National Technical Information Service (NTIS)

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-...

J. P. Adams M. L. Carboneau

1991-01-01

201

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

NASA Astrophysics Data System (ADS)

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

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

2013-04-01

202

Gamma-Ray Spectra of the Products of Fast Neutron Fission of U235 and U238 at Selected Times after Fission.  

National Technical Information Service (NTIS)

Experimental measurements of the gamma-ray pulse height distributions due to the products of fast-neutron-induced fission of U235 and U238 are presented. The measurements were made at nine selected times after fission from 15 minutes to 3 days. Irradiatio...

L. R. Bunney D. Sam

1966-01-01

203

RARE-EARTH METAL FISSION PRODUCTS FROM LIQUID U-Bi  

DOEpatents

Fission product metals can be removed from solution in liquid bismuth without removal of an appreciable quantity of uranium by contacting the liquid metal solution with fused halides, as for example, the halides of sodium, potassium, and lithium and by adding to the contacted phases a quantity of a halide which is unstable relative to the halides of the fission products, a specific unstable halide being MgCl/sub 3/.

Wiswall, R.H.

1960-05-10

204

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

Microsoft Academic Search

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

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

1979-01-01

205

Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations  

Microsoft Academic Search

The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF\\/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD\\/NEA burnup credit criticality benchmarks (Phase 1B) and

Do Heon Kim; Choong-Sup Gil; Jonghwa Chang; Yong-Deok Lee

2005-01-01

206

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

NASA Astrophysics Data System (ADS)

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

Beddingfield, David Harris

207

Fission product behavior during the in-pile Severe Fuel Damage Test SFD 1-4  

SciTech Connect

Fission product behavior during Severe Fuel Damage (SFD) Test SFD 1-4 is discussed in this paper. The SFD 1-4 fuel bundle consisted of 26 previously irradiated (36000 MWd/tU) and 2 fresh pressurized-water-reactor fuel rods, and 4 stainless-steel clad Ag-In-Cd control rods. The 1-m bundle was heated to temperatures in excess of 2400 K during the test to reproduce temperature, fuel damage, fission product release, hydrogen generation, and control rod material behavior that are representative of a small break loss of coolant accident in a light-water-reactor. Samples of the effluent exiting the damaged fuel bundle consisting of superheated steam, fission products, aerosols, and noncondensable gases were collected and analyzed. Measurements of deposition on sample system surfaces were also made. The samples were analyzed using a variety of analytical techniques to provide an understanding of the release, transport and deposition of fission products. The results are presented in terms of release fractions, release rates and the distribution of fission products. The probable chemical forms of the predominant fission product species are inferred from the solubility, transport and deposition behavior of key isotopes. 9 refs.

Vinjamuri, K.; Osetek, D.J.; Petti, D.A.; Meikrantz, D.H.

1988-01-01

208

Spontaneous-fission properties and production of heavy-element isotopes  

SciTech Connect

Spontaneous fission was discovered in /sup 238/U as a natural mode of decay as long ago as 1940. However, because of the long spontaneous-fission half-life of /sup 238/U of about 10/sup 16/ years, the decay rate (specific activity) was so low that detailed studies of the properties of spontaneous fission had to await the synthesis of isotopes of higher Z elements with shorter spontaneous-fission half-lives. During the 1960s, milligram quantities of /sup 252/Cf became available to researchers through the Transplutonium Production Program of the US Atomic Energy Commission. The availability of /sup 252/Cf with a spontaneous-fission half-life of 85 years, and its resultant high specific activity, stimulated many pioneering studies of the spontaneous-fission process. detailed measurements of the mass, charge, and kinetic-energy distributions of the fission fragments, of prompt neutron and photon emission from the fragments, and the interrelationship of these properties were reported by numerous investigators. Since that time, studies of the spontaneous-fission properties of many other isotopes, some with half-lives of less than a second, have been made. 36 references, 12 figures, 2 tables.

Hoffman, D.C.

1984-07-01

209

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

SciTech Connect

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

Deforest, D.L.

1991-12-01

210

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

211

Fission Gas Release Behavior from High Burnup UO2 Fuels under Rapid Heating Conditions  

Microsoft Academic Search

Fission gas release (FGR) behavior under rapid heating conditions of high burnup UO2 fuels with developed rim structure has been examined using two different out-of-pile heating techniques with no restraint pressure. The burnups of the fuel specimens were 3686 GWd\\/tU. The bare fuel specimens were heated up to 600-1,800C at heating rates of 1.7 to 4,600C\\/s. The FGR process strongly

Katsumi UNE; Shinji KASHIBE; Akira TAKAGI

2006-01-01

212

Mass-resolved angular distribution of fission products in the 20Ne+232Th reaction  

NASA Astrophysics Data System (ADS)

Mass-resolved angular distributions of fission product were measured in the 20Ne + 232Th reaction at Elab = 125.6 and 142.5 MeV using the recoil catcher technique followed by offline ?-ray spectrometry. Angular anisotropy was found to decrease with increasing asymmetry of mass division. Angular anisotropies of the fission products in the symmetric region were significantly higher compared to those calculated using the statistical saddle-point model. Experimental anisotropies could be explained after considering the contribution from pre-equilibrium fission. Use of barrier energies corresponding to different mass asymmetry values in the calculations could reasonably reproduce the mass dependence of angular anisotropies. The role of barrier energies in governing the angular anisotropy indicates that the mass dependence of anisotropy may possibly be a distinguishing feature of pre-equilibrium fission from quasifission, in which the composite system escapes into the exit channel without being captured inside the saddle point.

Tripathi, R.; Sodaye, S.; Sudarshan, K.; Guin, R.

2013-08-01

213

Analysis of the Fission Yeast Rad3+ Gene Product.  

National Technical Information Service (NTIS)

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

C. R. Chapman T. Enoch

2000-01-01

214

Iodine as Fission Product During a PWR Accident.  

National Technical Information Service (NTIS)

In the accident caused release of radioiodide, the radioactivity of I exp 131 is determining in the first phase. The fission iodine is mainly present as CsI or other iodides in the fuel element of a PWR. CsI represents the most thermodynamically stable fo...

B. Bendick

1982-01-01

215

Design of pellet surface grooves for fission gas plenum  

Microsoft Academic Search

In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the

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

1986-01-01

216

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

Microsoft Academic Search

Prompt gamma-ray cascades in N=83 fission product nuclei near 132Sn have been studied at Eurogam II using a 248Cm source. Cross coincidences observed between gamma rays from complementary light and heavy fission fragments were vital for isotopic assignments. Yrast states in the N=83 isotones 134Sb, 135Te, and 136I are reported. The interpretation of the level schemes is based mainly on

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

1997-01-01

217

Application of evaluated fission-product delayed-neutron precursor data in reactor kinetics calculations  

SciTech Connect

Evaluated fission-product yield and decay data have been used to describe 105 delayed neutron precursors explicitly in point reactor kinetics calculations. Results calculated for /sup 235/U thermal fission show that rod-drop reactivity values obtained from kinetics calculations with 6-group precursor data are considerably higher than those calculated with explicit delayed-neutron precursor data. The calculated kinetics associated with positive reactivity steps are significantly different.

Perry, R.T.; Wilson, W.B.; England, T.R.; Brady, M.C.

1985-01-01

218

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

SciTech Connect

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

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

1999-02-17

219

Fission-product-release signatures for LWR fuel rods failed during PCM and RIA transients  

SciTech Connect

This paper discusses fission product release from light-water-reactor-type fuel rods to the coolant loop during design basis accident tests. One of the tests was a power-cooling-mismatch test in which a single fuel rod was operated in film boiling beyond failure. Other tests discussed include reactivity initiated accident (RIA) tests, in which the fuel rods failed as a result of power bursts that produced radial-average peak fuel enthalpies ranging from 250 to 350 cal/g. One of the RIA tests used two previously irradiated fuel rods. On-line gamma spectroscopic measurements of short-lived fission products, and important aspects of fission product behavior observed during the tests, are discussed. Time-dependent release fractions for short-lived fission products are compared with release fractions suggested by: the Reactor Safety Study; NRC Regulatory Guides; and measurements from the Three Mile Island accident. Iodine behavior observed during the tests is discussed, and fuel powdering is identified as a source of particulate fission product activity, the latter of which is neglected for most accident analyses.

Osetek, D.J.; King, J.J.; Croucher, D.W.

1981-01-01

220

Fission Gas and Iodine Release Measured in IFA-430 Up to 15 GWd/T UO sub 2 Burnup.  

National Technical Information Service (NTIS)

The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Id...

A. D. Appelhans J. A. Turnbull R. J. White

1983-01-01

221

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

SciTech Connect

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

Crum, Jarrod V.; Vienna, John D.

2009-08-03

222

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

223

THE GAMMA-RAY SPECTROMETRY OF FISSION PRODUCTS. VI. THE CALCULATED AND EXPERIMENTAL GAMMA-RAY SCINTILLATION SPECTRA OF U²³⁵ FISSION PRODUCTS WITH CYLINDRICAL NaI (Tl) CRYSTALS  

Microsoft Academic Search

The calculation of scintillation spectra for the broad parallel gamma ; rays emitted from U²³⁵ fission products incident on a right circular ; cylindrical NaI (Tl) crystals is presented. Typical scintillation spectra from ; fission products are presented. The parameters of the spectra were tabulated as ; functions of time after fission. The calculated spectra with a 1.75-in.-dia x 1

Hattori

1961-01-01

224

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

225

FIPS: A Process for the Solidification of Fission Product Solutions Using a Drum Drier.  

National Technical Information Service (NTIS)

A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the c...

S. Halaszovich M. Laser E. Merz D. Thiele

1976-01-01

226

New human data versus estimates of effects of inhaling fission product mixtures.  

PubMed

Recently, data on exposures of humans as well as animals to fission products in plumes emitted by underground Soviet tests have been declassified by the Khazakhstan government and published in English. Similar human intakes of gross fission product mixtures that caused acute prodromal symptoms have not been previously reported. Animal experiments with such complex mixtures have not received sufficient support to provide data that could be reliably extrapolated with dose-response models to humans for use in triage of internally exposed persons. This commentary compares some of the acute prodromal effects on humans from the recently released Soviet data with the estimates of Cowan and Kuper, and later estimates by Brodsky and colleagues. The latter estimates are concluded to be safer, and more easily adaptable, for use in triage of persons exposed to internal deposition of fission products of various mixtures. PMID:19066480

Brodsky, Allen; Reeves, Glen

2009-01-01

227

PROCESS USING BISMUTH PHOSPHATE AS A CARRIER PRECIPITATE FOR FISSION PRODUCTS AND PLUTONIUM VALUES  

DOEpatents

A process is described for separating plutonium from fission products carried therewith when plutonium in the reduced oxidation state is removed from a nitric acid solution of irradiated uranium by means of bismuth phosphate as a carrier precipitate. The bismuth phosphate carrier precipitate is dissolved by treatment with nitric acid and the plutonium therein is oxidized to the hexavalent oxidation state by means of potassium dichromate. Separation of the plutonium from the fission products is accomplished by again precipitating bismuth phosphate and removing the precipitate which now carries the fission products and a small percentage of the plutonium present. The amount of plutonium carried in this last step may be minimized by addition of sodium fluoride, so as to make the solution 0.03N in NaF, prior to the oxidation and prccipitation step.

Finzel, T.G.

1959-03-10

228

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

229

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media  

NASA Astrophysics Data System (ADS)

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

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

2005-12-01

230

Natural gas production from Arctic gas hydrates  

Microsoft Academic Search

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

1993-01-01

231

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

232

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

SciTech Connect

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

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

1988-10-01

233

Gas Production in Reactor Materials  

SciTech Connect

This paper presents an overview of the principal nuclear reactions that are known to produce hydrogen and helium in irradiated materials and a summary of the comparison of measurements with predictions in various reactors. Hydrogen and helium are produced in all reactor materials by fast neutron reactions which typically have thresholds above 4 MeV. Selected elements also have thermal neutron gas production reactions that can be quite prolific, such as 6Li, 10B, and 14N, and there are a number of elements which produce transmutation products that have high thermal neutron gas production cross sections, most notably 59Ni produced by irradiation of Ni and 65Zn produced by irradiation of Cu or Zn. Since gas production cross sections are isotope-specific, gas production rates can change during irradiation due to transmutation effects or initial rates can be modified by isotopic tailoring of reactor materials.

Greenwood, Lawrence R.

2006-01-18

234

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.

235

Consistent theoretical model for the description of the neutron-rich fission product yields  

NASA Astrophysics Data System (ADS)

The consistent model for the description of the independent fission product formation cross-section at light projectile energies up to about 100MeV is described. Pre-compound nucleon emission is described in the framework of the two-component exciton model using the Monte Carlo method, which allows one to incorporate a time duration criterion for the pre-equilibrium stage of the reaction. The decay of the excited compound nuclei, formed after the pre-equilibrium neutron and proton emission, is treated within the time-dependent statistical model with the inclusion of the main dynamical effects of nuclear friction on the fission width and saddle-to-scission descent time. For each member of the compound nucleus ensemble at scission point, the primary fragment isobaric chain yields are calculated using the multimodal approach with the inclusion two superasymmetric fission modes. The charge distribution of the primary fragment isobaric chains was considered as a results of frozen quantal fluctuations of the isovector nuclear matter density at the finite scission neck radius. The calculated fission product formation cross-sections in the neutron, proton, and ? -rays induced fission of the heavy actinides are presented.

Rubchenya, V. A.; yst, J.

2012-04-01

236

DETECTION OF GASEOUS FISSION PRODUCTS IN WATER--A METHOD OF MONITORING FUEL SHEATHING FAILURES. (Revised Title)  

Microsoft Academic Search

The gaseous activities stripped from samples of effluent coolant from the NRU fuel elements tested in the central thimble of the NRX reactor (NRU loop) and from the NRX main effluent were investigated. The activities obtained from the NRU loop can be attributed to gaseous fission products oniy. Design data have been obtained for a ''Gaseous Fission Product Monitor'' to

P. R. Tunnicliffe; A. C. Whittier

1953-01-01

237

Permian basin gas production  

Microsoft Academic Search

Of the 242 major gas fields in the Permian basin, 67 are on the Central Basin Platform, 59 are in the Delaware basin, 44 are in the Midland basin, 28 are in the Val Verde basin, 24 are on the Eastern Shelf, 12 are in the Horshoe Atoll and eight are on the Northwest Shelf. Eleven fields have produced over

Haeberle

1995-01-01

238

Relative yields of U-235 fission products measured in a high level radioactive sludge at Savannah River Site  

SciTech Connect

This paper presents measurements of the concentrations of 42 of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at Savannah River Site. The 42 fision products make up 98% of the waste sludge. We used inductively coupled plasma-mass spectroscopy for the analysis. The relative yields for most of the fission products are in complete agreement with the known relative yields for the beta decay chains of the two asymmetric branches of the slow neutron fission of U-235. Disagreements can be reconciled based on the chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses. This paper presents measurements of the concentrations of 42 (98%) of the long-lived U-235 fission products in a high-level radioactive waste sludge stored at the Savannah River Site. We analyzed the sludge with inductively coupled plasma-mass spectroscopy. The relative yields for most of the fission products agree completely with the known relative vields for the beta decay chains of the two asymmetric: branches of the slow neutron fission of U-235. The chemistry of the fission products in the caustic waste sludges, the neutron fluences in SRS reactors, or interferences in the ICP-MS analyses explain the differences in the measured and calculated results.

Bibler, N.E.; Coleman, C.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Kinard, W.F. [Charleston Coll., SC (United States). Dept. of Chemistry

1992-10-01

239

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

Microsoft Academic Search

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

Adrienne M. LaFleur; William S. Charlton

240

THE SEPARATION AND DETERMINATION OF RUTHENIUM IN FISSION PRODUCTS BY LIQUID LIQUID EXTRACTION WITH PYRIDINE  

Microsoft Academic Search

A method of separating radioactive ruthenium from fission products and ; technetium in an aqueous solution containing sodium hydroxide and hypochlorite by ; liquid-liquid extraction with pyridine is discussed. In the process, ruthenium ; in various ionic forms and various oxidation states is oxidized to perruthenate ; with the hypochlorite, and the perruthenate is transferred into pyridine from the ;

Toshiyasu Kiba; Akiko Miura; Yasuyuki Sugioka

1963-01-01

241

The Linear Accelerator Fuel Enricher Regenerator (LAFER) and Fission Product Transmutor (APEX)  

Microsoft Academic Search

Two major problems face the nuclear industry today; first is the long-term supply of fissile material and second is the disposal of long-lived fission product waste. The high energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator can interact with a molten lead target to produce spallation and

Meyer Steinberg; James R. Powell; Hiroshi Takahashi; Pierre Grand; Herbert J. C. Kouts

1979-01-01

242

Measurement of actinide, fission and daughter product contamination in prairie soil  

Microsoft Academic Search

SummaryIn 1951, 6.7 liter of an aqueous solution of actinides, fission and daughter products leaked from an underground storage tank into prairie soil. In 2001, soil samples were collected from the site to determine the degree and extent of the contamination. This paper reports on the continuation of complementary analytical techniques used to determine the concentrations of the contaminants in

W. S. Andrews; X. Wang; K. A. M. Creber

2005-01-01

243

Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges  

Microsoft Academic Search

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

Bibler

2005-01-01

244

Release of fission and activation products during light water reactor core meltdown  

Microsoft Academic Search

The most relevant open questions combined with activity release during hypothetical core meltdown accidents refer to the chemical behavior of the highly reactive elements iodine, cesium, and tellurium, to the release characteristics of the medium-volatile fission and activation products, to the properties of the resulting aerosol particles, and to various phenomena during steam explosion and melt\\/concrete interaction. To answer some

H. Albrecht; V. Matschoss; H. Wild

1979-01-01

245

SEPARATION OF FISSION PRODUCTS BY LIQUID-LIQUID EXTRACTION WITH CUPFERRON CHLOROFORM  

Microsoft Academic Search

To separate the radioactive strontium and cesium from solutions ; containing gross fission products and other non-radioactive metallic ions, ; cupferron-chloroform extraction was performed. By changing the acidity of the ; aqueous solution from l n hydrochloric acid to pH 4, andi by performing ; extractions in each stage, the strontium and the cesium remained in the aqueous ; phase,

Toshiyasu Kiba; Mitsue Kanetani

1958-01-01

246

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

National Technical Information Service (NTIS)

The authors propose an alternative approach to dispose of long-lived fission products (LLFPs) of type-II, such as Se-79, Tc-99, Pd-107, Sn-126, I-129, Cs-135, and long-lived radioactive Zr-93 into outer solar space. An escape velocity from the solar syste...

H. Takahashi X. Chen

1996-01-01

247

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

PubMed

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

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

2005-01-01

248

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

PubMed

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

Yurevich, Vladimir

2005-01-01

249

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

Microsoft Academic Search

In response to the accident at Three Mile Island Unit 2 (TMI-2), the United States Nuclear Regulatory Commission (USNRC) initiated a series of Severe Fuel Damage tests that were performed in the Power Burst Facility at the Idaho National Engineering Laboratory to obtain data necessary to understand (a) fission product release, transport, and deposition; (b) hydrogen generation; and (c) fuel\\/cladding

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

1987-01-01

250

SEPARATION OF FISSION PRODUCT VALUES FROM THE HEXAVALENT PLUTONIUM BY CARRIER PRECIPITATION  

DOEpatents

An improved precipitation of fission products on bismuth phosphate from an aqueous mineral acid solution also containing hexavalent plutonium by incorporating, prior to bismuth phosphate precipitation, from 0.05 to 2.5 grams/ liter of zirconium phosphate, niobium oxide. and/or lanthanum fluoride is described. The plutonium remains in solution.

Davies, T.H.

1959-12-15

251

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

Microsoft Academic Search

Since less than 1% of the naturally occurring world supply of ruthenium (Ru), rhodium (Rh), and palladium (Pd) is available from US sources, an alternative unexploited source is the Ru, Rh, and Pd created as fission products during the burnup of nuclear fuel. The Pacific Northwest Laboratory (PNL), is conducting a research program to develop a cost effective, waste management-compatible

Lautensleger

1983-01-01

252

Fission Product Behavior During the In-Pile Severe Fuel Damage Test SFD 1-4.  

National Technical Information Service (NTIS)

Fission product behavior during Severe Fuel Damage (SFD) Test SFD 1-4 is discussed in this paper. The SFD 1-4 fuel bundle consisted of 26 previously irradiated (36000 MWd/tU) and 2 fresh pressurized-water-reactor fuel rods, and 4 stainless-steel clad Ag-I...

K. Vinjamuri D. J. Osetek D. A. Petti D. H. Meikrantz

1988-01-01

253

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

Microsoft Academic Search

A methodology based on chemical thermodynamics has been developed to treat the transport of volatile fission products (FPs) through the core and the primary system. The FPs considered are cesium, iodine, tellurium, strontium, and ruthenium, which may pose the major biohazard in postulated severe accidents in light water reactors. The vapor transport of FPs depends on the volatilities of the

D. Cubicciotti; B. R. Sehgal

1984-01-01

254

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

SciTech Connect

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

Lautensleger, A.W.

1983-10-01

255

COPAR-FD. Release of Metallic Fission Products from Coated Nuclear Fuel Particles  

Microsoft Academic Search

COPAR-FD is used to calculate the release of metallic fission products from coated nuclear fuel particles, using a finite-difference solution of the governing partial differential equation. COPAR-FD interfaces with the TRAMP and TRAFIC codes for calculating transport in and release from graphite fuel blocks.

F. Tzung; M. Richards

1992-01-01

256

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

Microsoft Academic Search

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

P. G. Ellison; P. R. Monson; H. A. Mitchell

1990-01-01

257

Modeling of fission product release from HTR (high temperature reactor) fuel for risk analyses  

Microsoft Academic Search

The US and FRG have developed methodologies to determine the performance of and fission product release from TRISO-coated fuel particles under postulated accident conditions. The paper presents a qualitative and quantitative comparison of US and FRG models. The models are those used by General Atomics (GA) and by the German Nuclear Research Center at Juelich (KFA\\/ISF). A benchmark calculation was

J. Bolin; K. Verfondern; T. Dunn; M. Kania

1989-01-01

258

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

259

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

Microsoft Academic Search

This paper is devoted to the study of apatites and zirconia used as condit ioning materials of nuclear wastes. Among the long-lived fission products 129 I and 99 Tc release has to be evaluated. First, the stable isotopes 187 Re and 98Mo considered as chemical homologous of Tc were implanted in hydroxyapatit e. Then, cladding tube pieces (hulls) present a

N. Millard-Pinard; F. Brossard; N. Chevarier; A. Chevarier; D. Crusset; C. Gaillard; N. Moncoffre; K. Poulard

260

Inventories of Actinide and Fission Product Arisings in Spent Nuclear Fuel: Results Using the RICE Code.  

National Technical Information Service (NTIS)

The CEGB reactor inventory code RICE has been used to generate comprehensive inventories of actinide and fission products produced in reactor fuel during irradiation in Magnox, AGR and CFR systems. Revised neutron cross-section data for Magnox, AGR and CF...

J. H. Mairs S. Nair

1979-01-01

261

New ENDF/B-VII.0 Evaluations of Neutron Cross Sections for 32 Fission Products.  

National Technical Information Service (NTIS)

Neutron cross sections for fission products play important role not only in the design of extended burnup core and fast reactors, but also in the study of the backend fuel cycle and the criticality analysis of spent fuel. New evaluations in both the reson...

H. Kim M. Herman P. Oblozinsly S. F. Mughabghab Y. O. Lee

2007-01-01

262

Intercomparison of Recent Evaluations for the Capture Cross Sections of Some Fission-Product Nuclides.  

National Technical Information Service (NTIS)

For 19 important fission-product nuclides, a comparison is made between various recent evaluations of neutron capture cross sections in the energy range of 1 MeV to 10 MeV. The results of recent differential and integral measurements are combined in order...

H. Gruppelaar A. J. Janssen J. W. M. Dekker

1976-01-01

263

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

Microsoft Academic Search

A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform

Gohar

2000-01-01

264

Effects of Soluble Fission Products on Thermal Conductivities of Nuclear Fuel Pellets  

Microsoft Academic Search

Simulated high burnup UO2 and (U, Gd)O2 pellets, doped with soluble fission product elements (Sr, Zr, Y, La, Ce, Nd), were prepared. Pellet thermal diffusivities were measured by a laser flash method and their thermal conductivities were evaluated. Thermal conductivities decreased with an increase in the total amount of soluble elements at low temperatures, while they were almost independent of

Shinji ISHIMOTO; Mutsumi HIRAI; Kenichi ITO; Yoshiaki KOREI

1994-01-01

265

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

National Technical Information Service (NTIS)

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

H. Gruppelaar

1977-01-01

266

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

SciTech Connect

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

Raman, S.; Murphy, B.D.

1995-09-01

267

Isotopic production cross sections of fission residues in 197Au-on-proton collisions at 800 A MeV  

Microsoft Academic Search

Interactions of 197Au projectiles at 800 A MeV with protons leading to fission are investigated. We measured the production cross sections and velocities of all fission residues which are fully identified in atomic and mass number by using the in-flight separator FRS at GSI. The new data are compared with earlier measurements of the characteristics of fission in similar reactions.

J. Benlliure; P. Armbruster; M. Bernas; A. Boudard; J. P. Dufour; T. Enqvist; R. Legrain; S. Leray; B. Mustapha; F. Rejmund; K.-H. Schmidt; C. Stephan; L. Tassan-Got; C. Volant

2001-01-01

268

Synthetic gas production  

Microsoft Academic Search

Disclosed is a method and apparatus for producing synthetic gas by using a blast furnace as a gasifier. The furnace is charged in a conventional manner with particles of solid carbonaceous material such as normal, low grade or undersized coke together with slag-producing material, such as limestone, silica and\\/or basic oxygen furnace and\\/or open hearth furnace slag. Fluent fuel such

Wintrell

1979-01-01

269

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

270

PERCOLATION ON GRAIN BOUNDARY NETWORKS: APPLICATION TO FISSION GAS RELEASE IN NUCLEAR FUELS  

SciTech Connect

The percolation behavior of grain boundary networks is characterized in two- and three-dimensional lattices with circular macroscale cross-sections that correspond to nuclear fuel elements. The percolation of gas bubbles on grain boundaries, and the subsequent percolation of grain boundary networks is the primary mechanism of fission gas release from nuclear fuels. Both radial cracks and radial gradients in grain boundary property distributions are correlated with the fraction of grain boundaries vented to the free surfaces. Our results show that cracks surprisingly do not significantly increase the percolation of uniform grain boundary networks. However, for networks with radial gradients in boundary properties, the cracks can considerably raise the vented grain boundary content.

Paul C. Millett

2012-02-01

271

Production of synthesis gas  

SciTech Connect

Synthesis gas prepared from ash-containing carbonaceous fuel is passed to a water-containing contace zone wherein the ash is collected in a lower settling portion of the contact zone prior to passage through a valved passageway and thence to a lock hopper during a valve-open period, flow of said ash being augmented by intermittent positive flow of water during the time said valve is open, through said valve from said settling zone to said lock hopper whereby more positive flow of ash is obtained (when compared to that obtained by use of continuous flow of water from the lock hopper) with substantially decreased heat loss.

Jahnke, F. C.; Crouch, W. B.

1985-08-06

272

Developing a computer code simulating recoil-beta decay tagging of fission products  

NASA Astrophysics Data System (ADS)

Neutron-rich nuclei far from stability exhibit insightful structural patterns and deformations, and can be used for testing existing nuclear theories. However, experimental information on many such nuclei is lacking, because they are hard to produce directly or through fusion-evaporation reactions. One way to populate light neutron-rich nuclei is by exploiting the fission process. Since these nuclei are primarily beta emitters, a technique of recoil beta tagging can be employed. A gas-filled separator, such as SASSYER (WNSL) or BGS (LBNL), can be used to select a mass window of fission fragments, which will be implanted on a DSSD detector located at the focal plane of the separator. By selecting high beta-endpoint energies, characteristic for the nuclei of interest, decays at the DSSD can be correlated to emitted gamma rays for further spectroscopy studies. An essential step in planning this project would be the development of a computer simulation of count rates at the DSSD. The program works by reading in files of half-lives and fission yields and uses a step-by-step iteration process. The role of the code is two-fold: to help identify a suitable nucleus for study and to optimize a mass window for its highest count rate. To test the method, experiments are planned with a fission source, such as Cf-252, placed at the target position of a recoil separator. If successful, the technique could be extended to in-beam experiments.

Darakchieva, Berta

2008-10-01

273

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

Microsoft Academic Search

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

V. Jedinkov

1983-01-01

274

Thermal release of volatile fission products from irradiated nuclear fuel  

Microsoft Academic Search

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

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

2011-01-01

275

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

NASA Astrophysics Data System (ADS)

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

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

1998-08-01

276

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

SciTech Connect

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

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

1989-11-01

277

Nuclear Fission  

NSDL National Science Digital Library

Start a chain reaction, or introduce non-radioactive isotopes to prevent one. Control energy production in a nuclear reactor! (Previously part of the Nuclear Physics simulation - now there are separate Alpha Decay and Nuclear Fission sims.)

Simulations, Phet I.; Adams, Wendy; Blanco, John; Lemaster, Ron; Mckagan, Sam; Perkins, Kathy

2004-07-01

278

Background for Terrestrial Antineutrino Investigations: Radionuclide Distribution, Georeactor Fission Events, and Boundary Conditions on Fission Power Production  

Microsoft Academic Search

Estimated masses of fissioning and non-fissioning radioactive elements and\\u000atheir respective distributions within the Earth are presented, based upon the\\u000afundamental identity of the components of the interior 82% of the Earth, the\\u000aendo-Earth, with corresponding components of the Abee enstatite chondrite\\u000ameteorite. Within limits of existing data, the following generalizations\\u000aconcerning the endo-Earth radionuclides can be made: (1) Most

J. Marvin Herndon; Dennis A. Edgerley

2005-01-01

279

Nonequilibrium behavior of fission gas bubbles with emphasis on the effects of the equation of state. [LMFBR  

Microsoft Academic Search

The paper presents a computer code designed to estimate fission gas behavior during transient fuel conditions, allowing for nonequilibrium bubble states, with emphasis on equation of state sensitivity. The computer code is a modification of the original code by R. G. Esteves, A. R. Wazzan, and D. Okrent, which in its present form includes the following: resolution, coalescence, leakage to

Steele

1976-01-01

280

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

SciTech Connect

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

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

1980-03-01

281

Flue gas desulfurization increasing productivity  

Microsoft Academic Search

This paper presents an analysis of those components of a wet flue gas desulfurization system which can affect the productivity of an electrical power generation plant. These complex systems are very important to an electrical power plant in desulfurizing flue gases in order to comply with Federal and State pollution control regulations. They must approach 100% availability so as not

Zourides

1983-01-01

282

EIA's Natural Gas Production Data  

EIA Publications

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

Information Center

2009-04-09

283

Electricity and Heat Production Using Natural Gas.  

National Technical Information Service (NTIS)

The purpose of this report is to describe technique, production costs and competitiveness for the production of electricity and heat from natural gas. The report deals with the production of electricity using gas turbines, conventional power plants fuelle...

E. Hakkarainen B. Olsson M. Borchers

1987-01-01

284

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

SciTech Connect

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

Lawrence, G.P.

1991-01-01

285

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

SciTech Connect

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

Robin D. Rogers

2004-12-09

286

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

SciTech Connect

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

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

2007-04-22

287

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

NASA Astrophysics Data System (ADS)

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

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

2012-08-01

288

Sensibility Analysis of the Effect of Various Key Parameters on Fission Product Concentration Mass Number 127 to 132 and Xe - 133 M).  

National Technical Information Service (NTIS)

An analytical sensitivity analysis has been made of the effect' of various parameters on the evaluation of fission product concentration. Such parameters include cross sections, decay constants, branching ratios, fission yields, flux and time. The formula...

A. Sola

1978-01-01

289

Calculation of the delayed gamma-ray energy spectra from aggregate fission product nuclides  

Microsoft Academic Search

The beta-delayed emission process of gamma rays was treated with a gross theory of beta decay and a cascade gamma transition model. The method proposed was applied to calculations of the delayed gamma-ray energy spectra for short-lived fission product nuclides that lack experimental information on their gamma-ray transition properties. The calculated results are used to complement the summation calculation of

T. Yoshida; J. Katakura

1986-01-01

290

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

Microsoft Academic Search

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

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

1958-01-01

291

The spatial distribution of fission product gamma-ray energy in reactor fuel elements  

Microsoft Academic Search

Thermoluminescent dosimeters were used to determine the spatial distribution of fission product gamma-ray energy deposition in pressurized water reactor (PWR) fuel under conditions similar to those of a postulated loss-of-coolant accident (LOCA). Measurements were made in a mock-up which simulated exactly the materials and dimensions of a PWR fuel element. The detectors were positioned at various positions in the UO

Bass; R. B. Jr

1978-01-01

292

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

Microsoft Academic Search

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

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

1973-01-01

293

SELECTIVE SEPARATION OF URANIUM FROM THORIUM, PROTACTINIUM AND FISSION PRODUCTS BY PEROXIDE DISSOLUTION METHOD  

DOEpatents

A method is described for separating U/sup 233/ from thorium and fission products. The separation is effected by forming a thorium-nitric acid solution of about 3 pH, adding hydrogen peroxide to precipitate uranium and thorium peroxide, treating the peroxides with sodium hydroxide to selectively precipitate the uranium peroxide, and reacting the separated solution with nitric acid to re- precipitate the uranium peroxide.

Seaborg, G.T.; Gofman, J.W.; Stoughton, R.W.

1959-08-18

294

RELEASE OF FISSION PRODUCTS ON THE IN-PILE MELTING OR BURNING OF REACTOR FUELS  

Microsoft Academic Search

In-pile experiments are being conducted to study the release of fission products during simulated reactor accidents. Two types of experiments have been performed in the Oak Ridge Research Reactor to simulate reactor accidents in which fuel elements are destroyed by melting or burning. One type consisted of melting or vaporizing a miniature stainless-steel-clad UO2 fuel element in a helium atmosphere.

R. P. Shields; W. E. Browning; Jr. C. E. Miller; Jr. B. F. Roberts

295

Observation and Measurement of Se79 in SRS High-Level Tank Fission Product Waste  

Microsoft Academic Search

The authors report the first observation of confirmed Se-79 activity in Savannah River Site high level fission product waste. Se-79 was measured after a seven step chemical treatment to remove interfering activity from Cs-137, Sr-90, and plutonium at levels 105 times higher than the observed Se-79 content and to remove Tc-99 at levels 300 times higher than observed Se-79. Se-79

Dewberry

2000-01-01

296

Thermodynamic analysis of chemical states of fission products in uraniumzirconium hydride fuel  

Microsoft Academic Search

The chemical state of fission products (FPs) in U+ZrH1.60 fuel was studied from the thermodynamic point of view. Twenty most abundant FP elements were taken into account in the system of UZrHOFP in which oxygen is treated as an impurity. The Thermo-Calc computer code was used to calculate the equilibrium state of the multi-phase and multi-component system. This calculation shows

Jintao Huang; Bun Tsuchiya; Kenji Konashi; Michio Yamawaki

2001-01-01

297

Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements  

Microsoft Academic Search

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW\\/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding

Vadim V. Konyashov; Alexander M. Krasnov

2002-01-01

298

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

Microsoft Academic Search

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

W. Schenk; H. Nabielek

1991-01-01

299

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

300

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

301

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

302

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

Microsoft Academic Search

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

YongDeok Lee; Jonghwa Chang; Pavel Oblozinsky

2000-01-01

303

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

Microsoft Academic Search

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

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

2010-01-01

304

PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION  

DOEpatents

An aqueous nitric acid solution containing uranium, plutonium, and fission product values is contacted with an organic extractant comprised of a trialkyl phosphate and an organic diluent. The relative amounts of trialkyl phosphate and uranium values are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 moles uranium per mole of trialkyl phosphate, thereby preferentially extracting uranium values into the organic extractant.

Ellison, C.V.; Runion, T.C.

1961-06-27

305

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

SciTech Connect

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

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

1988-03-01

306

The ?-ray spectrum of fission products from slow neutron irradiation of uranium 235  

Microsoft Academic Search

The fission products from uranium 235, irradiated by slow neutrons, have been analysed by a two-crystal ?-ray scintillation spectrometer. The spectrum between one day and seventy days after irradiation has been recorded. The gross spectrum varies with time and consists of two groups of lines around 0.2 and 0.7 MeV, with a single line at 1.58 MeV as the only

D H Peirson

1955-01-01

307

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

SciTech Connect

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

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

1990-12-31

308

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

SciTech Connect

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

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

1990-01-01

309

Fission product release and its environment impact for normal reactor operations and for relevant accidents  

Microsoft Academic Search

The radioactive concentration in the primary loop and the radioactive release for both normal operations and accidents for the HTR-10 are calculated and presented in the paper. The coated-particle fuel is used in the HTR-10, which has good performance of retaining fission products. Therefore the radioactive concentration in the primary loop of the HTR-10 is very low, and the amount

Liu Yuanzhong; Cao Jianzhu

2002-01-01

310

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

Microsoft Academic Search

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

J. P. Adams; M. L. Carboneau

1991-01-01

311

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

Microsoft Academic Search

Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000°C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released

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

1984-01-01

312

An analytical solution for simulation of the fission gas behaviors with time-dependent piece-wise boundary resolution  

NASA Astrophysics Data System (ADS)

An analytical solution of gas concentration for the equivalent spherical grain is obtained first in Laplace space, then the inverse-Laplace transformed solution is further developed. The corresponding analytical expressions for the grain boundary gaseous swelling and the fission gas release in UO2 nuclear fuels are developed in the absence of grain growth. The following phenomena and assumptions are taken into account in our model, including the gas atom diffusion, saturation and the time-varying piece-wise inter-granular resolution. The explicit expression for saturation time of the grain boundary gas atoms is also obtained. Our approximated analytical solutions for the fission gas behaviors are validated through comparison with those solved by finite difference method. Good agreement has been achieved for the cases with different input parameters. Based on the developed analytical solutions, the effects of the grain sizes and the external pressure on the fission gas behaviors are investigated. This study lays a foundation for the multi-scale simulation of the thermo-mechanical behaviors in nuclear fuel elements.

Cui, Yi; Huo, Yongzhong; Ding, Shurong; Zhang, Lin; Li, Yuanming

2012-05-01

313

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

Code of Federal Regulations, 2010 CFR

...production, by final product sold, of oil, gas, and other products. Disclosure...including transfers) per unit of oil, gas and other products produced; and...in common units of production with oil, gas, and other products converted...

2010-04-01

314

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

Code of Federal Regulations, 2010 CFR

...production, by final product sold, of oil, gas, and other products. Disclosure...including transfers) per unit of oil, gas and other products produced; and...in common units of production with oil, gas, and other products converted...

2009-04-01

315

Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations  

NASA Astrophysics Data System (ADS)

The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok

2005-05-01

316

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

SciTech Connect

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

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

1982-06-01

317

Fission product release as a function of chemistry and fuel morphology  

SciTech Connect

Analysis of the consequences of severe reactor accidents requires knowledge of the location and chemical form of fission products throughout the accident sequence. Two factors that strongly influence the location and chemical form of fission products are the chemistry within the core and the morphology of the fuel or fuel-bearing debris. This paper reviews the current understanding of the these factors garnered from integral and separate effect experiments and the TMI-2 accident, and provides perspective on the significance of contributing phenomena for the analysis of severe accidents, particularly during the in-vessel phase. Information has been obtained recently on phenomena affecting the release of fission products from fuel and the reactor vessel during the in-vessel melt progression phase of a severe accident. The influence of a number of these phenomena will be reviewed, including fuel chemistry, H{sub 2}/H{sub 2}O ratio, fuel liquefaction, molten pools, and debris beds. 13 refs., 1 fig., 1 tab.

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

1989-01-01

318

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

319

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

320

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

321

Production of reactive oxygen species in response to replication stress and inappropriate mitosis in fission yeast  

PubMed Central

Summary Previous studies have indicated that replication stress can trigger apoptosis-like cell death, accompanied (where tested) by production of reactive oxygen species (ROS), in mammalian cells and budding yeast (Saccharomyces cerevisiae). In mammalian cells, inappropriate entry into mitosis also leads to cell death. Here we report similar responses in fission yeast (Schizosaccharomyces pombe). We used ROS- and death-specific fluorescent stains to measure the effects of mutations in replication initiation and checkpoint genes in fission yeast on the frequencies of ROS production and cell death. We found that certain mutant alleles of each of the four tested replication initiation genes caused elevated ROS and cell death. Where tested, these effects were not enhanced by checkpoint gene mutations. Instead, when cells that were competent for replication but defective in both the replication and damage checkpoints were treated with hydroxyurea, which slows replication fork movement, the frequencies of ROS production and cell death were greatly increased. This was a consequence of elevated CDK activity, which permitted inappropriate entry into mitosis. Thus studies in fission yeast are likely to prove helpful in understanding the pathways that lead both from replication stress and from inappropriate mitosis to cell death in mammalian cells.

Marchetti, Maria A; Weinberger, Martin; Murakami, Yota; Burhans, William C; Huberman, Joel A

2006-01-01

322

Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations  

SciTech Connect

The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok [Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon, 305-600 (Korea, Republic of)

2005-05-24

323

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.

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

2013-07-01

324

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

325

Chemistry of Sodium Coolant. Impurities of Nuclear Fuel and Its Fission Products in Sodium Coolant of Fast Reactors.  

National Technical Information Service (NTIS)

Properties of impurities of nuclear fuel and its fission products and their behaviour in sodium coolant of the primary circuit of fast reactors are considered. The impurities are classified with account of their properties, behaviour in the circuit and si...

E. E. Konovalov A. I. Lastov P. S. Otstavnov

1982-01-01

326

ESOL (Elemental Separation On Line) Facility for the Generation and Radiochemical Separation of Short Half-Life Fission Products.  

National Technical Information Service (NTIS)

A facility has been developed at the Idaho National Engineering Laboratory (INEL) for the generation and rapid radiochemical separation of short half-life mixed fission products. This facility, referred to as the Idaho Elemental Separation On Line (ESOL),...

R. J. Gehrke D. H. Meikrantz J. D. Baker R. A. Anderl V. J. Novick

1988-01-01

327

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

National Technical Information Service (NTIS)

This project has demonstrated that Sr2+ and Cs+ can be selectively extracted from aqueous solutions into ionic liquids using crown ethers and that unprecedented large distribution coefficients can be achieved for these fission products. The volume of seco...

2004-01-01

328

Experience in Monitoring the BWR (Boiling Water Reactor) Fuel Behaviour and Fission Product Releases During off Normal Conditions.  

National Technical Information Service (NTIS)

Tarapur Atomic Power Station has accumulated over 33 reactor years of operating experience in monitoring Boiling Water Reactor fuel behavior. The sudden and sharp increases in the fission product releases were experienced in the earlier years due to gross...

N. B. Joshi V. S. inivasan K. Nanjundeswaran

1986-01-01

329

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

PubMed

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

Mushtaq, A

2011-01-08

330

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

Microsoft Academic Search

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

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

2006-01-01

331

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

332

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

Microsoft Academic Search

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

Kouki MORIYAMA; Hirotaka FURUYA

1997-01-01

333

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

NASA Astrophysics Data System (ADS)

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

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

1996-10-01

334

Systematic Analysis of Structural Effects in Fission-Product Yields and Neutron Data and the Consequences for Our Understanding of the Fission Process and the Predictive Power of Model Predictions  

NASA Astrophysics Data System (ADS)

Structural effects in fission-product yields and neutron data for a large number of fissioning nuclei between 220Th and 256Fm from spontaneous fission to 14-MeV-neutron-induced fission have been used to deduce information on the properties of the fissioning systems. Macroscopic properties are attributed to the compound nucleus, while fission channels are ascribed to shells in the nascent fragments. Using a recent general empirical description of the nuclear level density and assuming different characteristic time scales for the collective degrees of freedom of the fissioning system, a new fission model has been developed. The model combines the statistical concept of the scission-point model of Wilkins et al. with empirically determined properties of the potential-energy surface and some characteristic dynamical freeze-out times. Although no fine tuning of the parameters has yet been performed, the model reproduces all measured fission yields and neutron data rather well with a unique set and a relatively small number of free parameters. Since the parameters of the model are closely related to physical properties of the systems, some interesting conclusions on the fission process can be deduced. Prospects for the predictive power of this semi-empirical approach for hitherto unknown fissioning systems are discussed.

Schmidt, K.-H.; Jurado, B.

2011-10-01

335

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

NASA Astrophysics Data System (ADS)

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

Juhasz, Albert J.

336

On the Correction Factors Affecting the Evaluation of the Beta Total Radioactivity of Fission Products; SUR LES FACTEURS DE CORRECTION AFFECTANT L'EVALUATION DE LA RADIOACTIVITE BETA TOTALE DES PRODUITS DE FISSION  

Microsoft Academic Search

Data are presented which make possible the precise evaluation of the ; BETA activity of fission products measured in a variety of conditions of ; absorption and self-absorption. The correction factors to apply to the BETA ; measurement of fission products when the counter has been standardized wlth ; potassium-40 or with thallium-204, are tabulated. Nomograms for computing self-; absorption

P. Gaglione; E. Van der Stricht

1963-01-01

337

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

Microsoft Academic Search

The discovery of naturally occurring fission reactors in the rock strata of the Paleoproterozoic Francevillian Basin in the Republic of Gabon in equatorial West Africa led to several programs to define migration and\\/or retention of uranium and fissiogenic isotopes from\\/in the natural reactor zones. Although much understanding has been gained, new insight is needed regarding the chemical and physical parameters

B. Nagy; M. J. Rigali

1993-01-01

338

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

339

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

340

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

341

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

NASA Astrophysics Data System (ADS)

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

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

2008-10-01

342

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

Microsoft Academic Search

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

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

1978-01-01

343

Isomer production ratios and the angular momentum distribution of fission fragments  

NASA Astrophysics Data System (ADS)

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

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

2013-10-01

344

Nuclear Fission  

Microsoft Academic Search

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

I. Halpern

1959-01-01

345

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

NASA Astrophysics Data System (ADS)

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

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

1993-07-01

346

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

Microsoft Academic Search

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

C. F. Ksanda; E. Lauments

1959-01-01

347

Angular distributions of specific gamma rays emitted in the deexcitation of prompt fission products of 252Cf  

Microsoft Academic Search

Angular distributions of specific gamma rays emitted in the deexcitation of prompt fission products of 252Cf were measured with respect to the fission direction. A total of 42 angular distributions were measured, 23 of which were of transitions in even-even fragments. The strong anisotropy (A2=0.4-0.6) measured for 2+--> 0+, 4+--> 2+, and 6+ --> 4+ transitions in 138,140Xe and 142,144Ba

A. Wolf; E. Cheifetz

1976-01-01

348

Determination of the antineutrino spectrum from 235U thermal neutron fission products up to 9.5 MeV  

Microsoft Academic Search

A measurement of the cumulated beta spectrum from 235U fission products was performed with a magnetic spectrometer. The energy range 2.0-9.5 MeV was covered achieving a sensitivity of 3 x 10-6 betas per MeV per fission in the high energy region. The absolute rates were calibrated with a precision of 3%. The beta spectrum was converted into the correlated nue

K. Schreckenbach; G. Colvin; W. Gelletly; F. von Feilitzsch

1985-01-01

349

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

PubMed

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

Shozugawa, Katsumi; Nogawa, Norio; Matsuo, Motoyuki

2012-01-20

350

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

Microsoft Academic Search

The Californium Rare Ion Breeder Upgrade (CARIBU) for the ATLAS facility is under construction. The facility will use 252Cf fission fragments thermalized and collected into a low-energy particle beam by a helium gas catcher. In order to reaccelerate these beams an existing ATLAS ECR ion source is being reconfigured as a charge breeder source. A 1Ci 252Cf source will provide

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

2007-01-01

351

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

352

Level Structure of 138Cs from the Short-Lived Fission Product 138Xe  

Microsoft Academic Search

Gamma rays from the 15-min fission product 138Xe were studied with Ge(Li) detectors, NaI(Tl) crystal and coincidence equipments. Gamma rays of the following energies and relative intensities have been observed: 153.2 (26), 242.8 (9.6), 258.2 (100), 396.5 (16), 401.7 (9.0), 434.6 (48), 1770 (39), 2002 (14) and 2013 keV (29). With the measurements of the single gamma ray spectrum and

Teruaki Nagahara; Kenji Tomura; Nobuyoshi Miyaji; Hideaki Kurihara; Yukiko Mizuno

1969-01-01

353

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

SciTech Connect

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

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

1997-01-01

354

Fission product source terms for the LWR loss-of-coolant accident  

SciTech Connect

Models for cesium and iodine release from light-water reactor (LWR) fuel rods failed in steam were formulated based on experimental fission product release data from several types of failed LWR fuel rods. The models were applied to a pressurized water reactor (PWR) undergoing a hypothetical loss-of-coolant accident (LOCA) temperature transient. Calculated total iodine and cesium releases from the fuel rods were 0.053 and 0.025% of the total reactor inventories of these elements, respectively, with most of the release occurring at the time of rupture. These values are approximately two orders of magnitude less than releases used in WASH-1400, the Reactor Safety Study.

Lorenz, R.A.; Collins, J.L.; Malinauskas, A.P.

1980-07-01

355

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

SciTech Connect

Best-estimate calculations of realistic source terms are presented which reduce uncertainties in predicting fission product release from the UO{sub 2} fuel over the temperature range between 770 K and 3000 K. The proposed method of correlation includes such fuel morphology effects as equiaxed fuel grain growth and fuel-cladding interaction. The method correlates the product of fuel release rate and equiaxed grain size with the inverse fuel temperature to yield a bulk mass transfer correlation. It was found that less and slower releases are predicted utilizing the bulk mass transfer correlation than such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation. A Severe Fuel Damage (SFD) analysis code was developed to perform the thermal hydraulic and fission product calculations needed to analyze the Power Burst Facility SFD tests. The predictions utilizing the bulk mass transfer correlations overall followed the experimental time-release histories during the course of the heatup, power hold and cooldown phases of the transients. Good agreements were achieved for the integral releases both in timing and in magnitude. The proposed bulk mass transfer correlations can be applied to both current and advanced light water reactor fuels. 17 refs., 8 figs., 3 tabs.

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

1989-05-01

356

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

357

Separation of fission products from spent pressurized water reactor fuels by anion exchange and extraction chromatography for inductively coupled plasma atomic emission spectrometric analysis  

Microsoft Academic Search

A study has been carried out on the separation of fission products from spent pressurized water reactor (PWR) fuels and their quantitative determination using inductively coupled plasma atomic emission spectrometry (ICP-AES). Plutonium and uranium were separated from fission products using anion exchange and tri-n-butylphosphate (TBP) extraction chromatography, respectively. Americium was separated and fission products such as Sr, Ba, Cd, La,

Chang Heon Lee; Moo Yul Suh; Kwang Soon Choi; Jung Suk Kim; Byong Chul Song; Kwang Yong Jee; Won Ho Kim

2001-01-01

358

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

NASA Astrophysics Data System (ADS)

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

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

2004-08-01

359

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

SciTech Connect

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

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

1986-05-01

360

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

NASA Astrophysics Data System (ADS)

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

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

2013-06-01

361

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

SciTech Connect

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

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

1996-10-01

362

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

363

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

364

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

365

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

Microsoft Academic Search

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

Hideaki Kuraishi; Tetsuo Sawada; Hisashi Ninokata; Hiroshi Endo

2001-01-01

366

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

NASA Astrophysics Data System (ADS)

To help understand the mechanisms of irradiation-induced defect formation and evolution in nuclear fuel, systematic experimental efforts have been carried out. Ceria (CeO2) was selected as a surrogate material for Uranium Dioxide (UO2) due to many similar properties. Lanthanum (La) was chosen as a dopant in CeO2 to investigate the effect of impurities. The presence of La in the CeO2 lattice introduces a predictable initial concentration of oxygen vacancies, making it possible to characterize hypostoichiometric effects in CeO2. The influence of two La concentrations, 5% and 25%, were examined.In situ Transmission Electron Microscopy (TEM) experiments were used to study the evolution of defect clusters and the influence of irradiation with two common fission products: Xe and Kr. The irradiations were performed on thin film, single crystal materials. The irradiation damage caused formation of dislocation loopsat 600 C and defect clusters at room temperature. Dislocation networks form as the result of interactions of defect clusters. The dislocation loops were determined to be mainly of 1/9[1 1 1] interstitial type loops. Quantitative results were obtained to characterize the fluence and temperature effects of irradiation. Slow defect kinetics were found with irradiation on 25% La doped CeO2 at 600 C and it is attributed to the higher concentration of oxygen vacancies due to high La dopant level.

Yun, D.; Ye, B.; Oaks, A. J.; Chen, W.; Kirk, M. A.; Rest, J.; Yacout, A. M.; Stubbins, J. F.

2012-02-01

367

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

368

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

NASA Astrophysics Data System (ADS)

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

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

2008-12-01

369

Nuclear Fission  

Microsoft Academic Search

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

ERICH SCHNEIDER; WILLIAM C. SAILOR

2006-01-01

370

Use of Gamma spectrometry for measuring fission product releases during a simulated PWR severe accident: Application to the VERDON experimental program  

Microsoft Academic Search

The release of fission products (FP) from a pressurized water reactor (PWR) during a hypothetical severe accident is a major topic in nuclear reactor safety assessment, since they are the main contributors to the source term in the environment. Fission products with short half lives are of particular importance due to their potential high radiological effects. In order to precisely

G. Ducros; S. Bernard; M. P. Ferroud-Plattet; O. Ichim

2009-01-01

371

Kema Scientific and Technical Reports, Volume 8, Number 1. Special Issue: Experiments on the High-Temperature Behaviour of Neutron-Irradiated Uranium Dioxide and Fission Products.  

National Technical Information Service (NTIS)

The objective of the study was to determine the release rate of fission products from overheated UO2, the chemical form of these fission products, and the transport mechanisms inside the nuclear fuel. UO2 spheres of approx. 1 mm diameter, irradiated in th...

R. H. J. Tanke

1990-01-01

372

Desulfurized gas production from vertical kiln pyrolysis  

Microsoft Academic Search

A gas, formed as a product of a pyrolysis of oil shale, is passed through hot, retorted shale (containing at least partially decomposed calcium or magnesium carbonate) to essentially eliminate sulfur contaminants in the gas. Specifically, a single chambered pyrolysis vessel, having a pyrolysis zone and a retorted shale gas into the bottom of the retorted shale zone and cleaned

Harry A. Harris; Jones Jr. John B

1978-01-01

373

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

NASA Astrophysics Data System (ADS)

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

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

2007-06-01

374

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

SciTech Connect

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

Dewberry, R.A.

2000-08-21

375

Transuranic and fission product contamination in lake sediments from an alpine wetland, Boron (France).  

PubMed

Transuranics and fission products have been measured in lake sediment samples, collected in an alpine wetland, to determine their vertical distribution and calculate inventories. The radionuclides considered are 90Sr, 137Cs, 238Pu, 239/240Pu and 241Am. From the results, a better knowledge of radionuclide accumulation mode and behaviour was obtained. In addition, the origins of the individual pollutants could be deduced from activity ratios. Analyses were made on different sediment cores. The sampling sites were chosen to enable future determination of the mass balances of the radiopollutants. As the selected study area is in a recreational area used by urban populations, a rough estimate was made of the mean external dose from 137Cs for comparison with the French regulation. PMID:16150519

Schertz, M; Michel, H; Barci-Funel, G; Barci, V

2005-09-16

376

Accident management to prevent containment failure and reduce fission product release  

SciTech Connect

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

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

1991-01-01

377

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

378

Passive nondestructive burnup monitoring of MNSR irradiated fuel by measuring photoneutrons produced within fission products.  

PubMed

A passive nondestructive method for monitoring of Syrian miniature neutron source reactor (MNSR) fuel burnup is introduced. The inner irradiation site design inside the Be reflector was exploited to measure the generated photoneutrons induced by fission products hard gamma radiation in the subcritical state. The photoneutron flux was measured using gold foils as a function of cooling time and operation power. For cooling time ranges between 10 and 25d, experiments show that (140)Ba is the extremely dominating inducer of photoneutrons and the measured flux is proportional to the accumulated (140)Ba. This result forms a new method base for MNSR fuel burnup monitoring. It might be used also as a safeguards technique to check the operator declared information. PMID:19620012

Haddad, Kh

2009-06-26

379

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

PubMed

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

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

2011-07-26

380

Magnitude of fission product depositions from atmospheric nuclear weapon test fallout in France.  

PubMed

The external dose attributable to fallout from worldwide atmospheric nuclear testing, which represents about 40% of the total effective dose received before 2000, is dominated by specific fission products such as 95Zr, 104Ba, 106Ru, 103Ru, and 144Ce, which are far less well-documented than 90Sr and 137Cs. The depositions of these nuclides over France were calculated on the basis of activity measurements in air and rainwater samples collected from 1961 to 1977. These depositions were then compared to the same radionuclides activities measured in grass during that period. This study shows that the transfer and deposition processes occur in a very similar manner for all the studied radionuclides. Depositions calculated in this study, consistent in most cases with UNSCEAR estimates, constitute a good basis for the external dose assessment of nuclear weapon test fallout over Western Europe. PMID:15057055

Renaud, Philippe; Louvat, Didier

2004-04-01

381

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

382

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.

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

2005-01-01

383

Methane Hydrate Gas Production by Thermal Stimulation.  

National Technical Information Service (NTIS)

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

P. L. McGuire

1981-01-01

384

Fission product release assessment for O6 mixed bundle channel feeder stagnation break in CANDU-6 reactor.  

National Technical Information Service (NTIS)

A fission product release assessment for O6 'mixed-bundle' channel feeder stagnation break in CANDU-6 reactor has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The tot...

D. J. Oh D. Bowslaugh

1996-01-01

385

THE GAMMA-RAY SPECTROMETRY OF FISSION PRODUCTS. V. GAMMA-RAY SPECTROMETRIC ANALYSIS OF FALLOUT SAMPLES  

Microsoft Academic Search

The gamma spectrometric method for the analysis of fission products was ; applied to some fall-out samples. They included snow, dust in the open air, and ; ashes of plants. Comparison was made between the apparent ages of samples from ; the experimental gamma -ray spectra with those from the beta decay curves of ; the same samples. The agreement

Hattori

1961-01-01

386

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

Microsoft Academic Search

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors

Eduardo Henrique Rangel Honaiser

2003-01-01

387

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

National Technical Information Service (NTIS)

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

N. A. Hanan

1998-01-01

388

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

389

Merging the CEM2K and LAQGSM Codes with GEM2 to Describe Fission and Light-fragment Production  

Microsoft Academic Search

We present the current status of the improved Cascade-Exciton Model (CEM)\\u000acode CEM2k and of the Los Alamos version of the Quark-Gluon String Model code\\u000aLAQGSM. To describe fission and light-fragment (heavier than He4) production,\\u000aboth CEM2k and LAQGSM have been merged with the GEM2 code of Furihata. We\\u000apresent some results on proton- and deuteron-induced spallation, fission, and\\u000afragmentation

Stepan G. Mashnik; Konstantin K. Gudima; Arnold J. Sierk

2003-01-01

390

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

391

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

NASA Astrophysics Data System (ADS)

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

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

2012-10-01

392

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

Microsoft Academic Search

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

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

1980-01-01

393

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

NASA Astrophysics Data System (ADS)

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

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

2009-03-01

394

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

NASA Astrophysics Data System (ADS)

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

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

2012-05-01

395

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

396

Excitation of Nitrogen Gas by Fission Fragments along the Entire Track  

Microsoft Academic Search

Luminescence produced in nitrogen in small segments of the tracks of fission fragments was studied with a pulse technique. The amount of formation of the C3?u state of nitrogen per unit distance, L*, was measured over the entire path length of the low-mass fragment group as a function of track length. The energy loss per unit distance dE\\/dr can be

John T. Sears; Robert Rodgers

1967-01-01

397

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 State of Alaska and included to obtain a U.S. Total.

Information Center

2013-08-30

398

Lake Erie gas production posed unique problems  

SciTech Connect

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

Sangster, R.B.

1981-09-01

399

Oil and gas leasing\\/production program  

Microsoft Academic Search

As the Congress declared in the Outer Continental Shelf Lands Act the natural gas and oil production from the Outer Continental Shelf constitutes an important part of the Nation's domestic energy supply. Federal offshore minerals are administered within the Department of the Interior by the Minerals Management Service (MMS), which provides access to potential new sources of natural gas and

Heimberger

1992-01-01

400

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

401

Deposition and distribution of Chernobyl fallout fission products and actinides in a Russian soil profile.  

PubMed

In this article the distribution of fission products and actinides in a soil profile from Novo Bobovicky in Russia, which was contaminated due to the Chernobyl nuclear power plant accident, is described. The ground deposition of long-lived fission products determined by gamma-spectrometry was (recalculated to 26 April 1986) 1600 kBq (137)Cs/m(2), 900 kBq (134)Cs/m(2) and 60 kBq (125)Sb/m(2). Of these radionuclides (137)Cs shows the dominating activity at the present time. After 6.5 years 90% of the Cs and Sb activity was contained in the upper 4 cm. A (239,240)Pu ground deposition of 77.4+/-8.0 Bq/m(2) was determined by alpha-spectrometry. The (238)Pu/(239,240)Pu activity ratio of 0.30+/-0.03 and (241)Pu/(239,240)Pu activity ratio of 115+/-14 (in 1986) measured in the soil profile, indicates that the analysed Pu originates mainly from the Chernobyl accident. The average (234)U/(238)U activity ratio of 1.06+/-0.29 indicates that the uranium in this soil is dominated by naturally occurring uranium. The alpha- and beta-autoradiography revealed that the activity is mainly present in particulate form. It could further be observed that the spots containing alpha- or beta-activity originated from different particles. A comparison of alpha-autoradiography with the bulk Pu and Am activity showed that 92% of the alpha-activity was present as clearly detectable alpha-spots. The beta-active particles, located by beta-autoradiography were correlated with gamma-spectrometric measurements and contained only (137)Cs. These hot spots ranged from 0.02 to 0.15 Bq.It could be concluded that the vertical transport of (137)Cs and fuel fragments occurs mainly by movement of particles through the soil. It could also be concluded that the fuel fragments found, in this soil were depleted in respect to Cs, Sb and Eu. Comparison of the analysed (238)Pu/(239,240)Pu, (241)Pu/(239,240)Pu and (241)Am/(239,240)Pu ratios with the ratios calculated with ORIGEN-S code gave an estimate of the average burn-up of the fuel particles to be in the range of 11-12 GWd/tU. The results presented in this article are valid for this single soil profile and should not be generalised unless validated in a more rigorous study of a larger number of soil profiles. PMID:12726697

Carbol, P; Solatie, D; Erdmann, N; Nyln, T; Betti, M

2003-01-01

402

/sup 244/CmO/sub 2//nat. -UO/sub 2/ hybrid blanket with flat fission power production  

SciTech Connect

In the present study, /sup 244/CmO/sub 2/ is mixed with nat.-UO/sub 2/ for the purpose of power flattening in a hybrid blanket with a reasonably high energy multiplication factor. Also, the temporal variations of the fission power density (FPD) are observed during an 18-month plant operation period. The main conclusion drawn from this work is that it became possible to keep a flat fission power profile (FPP) over a very long plant operation period of 18 months by simply omitting the beryllium multiplier in the blanket and keeping the neutron spectrum fairly unchanged throughout the fission zone. This reduced the efforts for fuel management to a minimum. A further observation focused on only minor variations of the integral neutronic data over longer plant operation periods. Among others, the fission power generation increase is also very modest. This results in an optimum investment for the nonnuclear island. The blanket burns up high-level nuclear waste /sup 244/Cm effectively, with efficient electricity production and breeding of a new type of nuclear fuel /sup 245/Cm with very superior nuclear properties. Finally, a warning should be issued for the careful international safeguarding of such a hybrid plant due to the extremely high quality of the bred plutonium fuel.

Sahin, S.; Erisen, A.; Cebi, Y.

1988-01-01

403

Electrochemical separation of actinides and fission products in molten salt electrolyte  

SciTech Connect

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

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

1995-09-15

404

Migration energies of native defects and fission products in uranium dioxide  

NASA Astrophysics Data System (ADS)

Despite the importance of fission products like Xe in nuclear fuels, the mechanism of how these atoms diffuse in the lattice is not known. In an effort to identify this mechanism, we have used density functional theory as well as a variety of different classical potentials for to study the migration energies of a variety of atomic steps in UO2, with and without Xe impurities and native defects. We find that the classical potential of Basak gives results which compare favorably with density functional theory for the diffusion of a Schottky defect cluster. We observe a new path for xenon-tetravacancy (a UO2 Schottky defect plus an additional U vacancy) motion using molecular dynamics. This path has a lower energy barrier than previously reported xenon-tetravacancy paths. We examine the possibility of a uranium vacancy dissociating from the xenon-tetravacancy cluster and find that large barriers for this dissociation. We also calculate xenon-double Schottky defect migration and find it has a slightly larger barrier than xenon-tetravacancy motion with the oxygen vacancies being weakly bound to the defect.

Thompson, Alexander; Wolverton, Chris

2011-03-01

405

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

SciTech Connect

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

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

1993-09-01

406

Composition of high fission product wastes resulting from future reprocessing of commercial nuclear fuels  

SciTech Connect

Pacific Northwest Laboratory studies, aimed at defining appropriate glass compositions for future disposal of high-level wastes, have developed composition ranges for the waste that will likely result during reprocessing of Light Water Reactor (LWR) and Liquid Metal Reactor (LMR) fuels. The purpose of these studies was to provide baseline waste characterizations for possible future commercial high-level waste so that waste immobilization technologies (e.g., vitrification) can be studied. Ranges in waste composition are emphasized because the waste will vary with time as different fuels are reprocesses, because choice of process chemicals is nuclear, and because fuel burnups will vary. Consequently, composition ranges are based on trends in fuel reprocessing procedures and on achievable burnups in operating reactors. In addition to the fission product and actinide elements, which are the primary hazardous materials in the waste, likely composition ranges are given for inert elements that may be present in the waste. These other elements may be present because of being present in the fuel, because of being added as process chemical during reprocessing, because of being added during equipment decontamination, or because of corrosion of plant equipment and/or fuel element cladding. This report includes a discussion of the chemicals added in variation of the PUREX process, which is likely to remain the favored reprocessing technique for commercial nuclear fuels. Consideration is also given to a pyrochemical process proposed for the reprocessing of some LMR fuels.

Swanson, J.L

1986-07-01

407

Fission product iodine release and retention in nuclear reactor accidents experimental programme at PSI  

NASA Astrophysics Data System (ADS)

Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

2003-01-01

408

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

SciTech Connect

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

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

2004-10-01

409

A potential photo-transmutation of fission products triggered by Compton backscattering photons  

NASA Astrophysics Data System (ADS)

We investigated the transmutation of some fission product nuclides I129, Cs135, Sn126, Zr93, Pd107, Cs137 and Sr90, induced by the Compton backscattering (CBS) photons generated from the future Shanghai Laser Electron Gamma Source (SLEGS) facility. The evaluated photo-transmutation rates for I129, Cs135, Sn126, Zr93, Pd107, Cs137 and Sr90 can achieve 2.510, 1.310, 4.810, 2.710, 9.410, 1.310 and1.610 per second, respectively, improving 4-5 orders of magnitude compared with those via the bremsstrahlung photons by a 10W/cm laser. The maximum transmutation coupling efficiencies of the CBS photons were estimated to be 1.36% for I129, 1.70% for Cs135, 2.02% for Sn126, 1.03% for Zr90, 1.52% for Pd107, 1.62% for Cs137 and 1.72% for Sr90, which are 2-6 times as those via the bremsstrahlung method by the 10W/cm laser. Moreover, we presented a possible experimental method for the future SLEGS facility to check the estimated results.

Chen, J. G.; Xu, W.; Wang, H. W.; Guo, W.; Ma, Y. G.; Cai, X. Z.; Lu, G. C.; Xu, Y.; Pan, Q. Y.; Fan, G. T.; Shen, W. Q.

2009-02-01

410

Oil and Gas Leasing/Production Program  

SciTech Connect

As the Congress declared in the Outer Continental Shelf Lands Act (OCSLA), the natural gas and oil production from the Outer Continental Shelf (OCS) constitutes an important part of the Nation's domestic energy supply. Federal offshore minerals are administered within the Department of the Interior by the Minerals Management Service (MMS), which provides access to potential new sources of natural gas and oil offshore by conducting lease sales. Each year, on or before March 31, the MMS (as mandated by OCSLA) presents to Congress a fiscal year annual report on the Federal offshore natural gas and oil leasing and production program. In FY 1990, the MMS's offshore natural gas and oil leasing and production program was the fourth largest producer of revenue for the US Treasury, contributing more than $3.0 billion. This report describes sales, exploration activities, and environmental monitoring activities. 16 figs., 11 tabs.

Heimberger, M.L.; O'Brien, D. (comps.)

1991-03-31

411

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

PubMed

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

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

2012-02-18

412

Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel  

SciTech Connect

Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance.

McDeavitt, S.M.; Abraham, D.P.; Park, J.-Y. [Argonne National Lab., IL (United States); Keiser, D.D. Jr. [Argonne National Lab., Idaho Falls, ID (United States)

1996-07-01

413

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

SciTech Connect

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

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

2012-01-01

414

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

Microsoft Academic Search

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

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

2008-01-01

415

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

NASA Astrophysics Data System (ADS)

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

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

2008-07-01

416

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

Microsoft Academic Search

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

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

2008-01-01

417

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

418

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

Microsoft Academic Search

Experimental measurements of delayed fission-product gamma-ray transmission through low-enriched UO fuel pin lattices in an air medium were conducted at the Rensselaer Polytechnic Institute Reactor Critical Facility (RCF). The RCF core consists of excess Special Power Excursion Reactor Test (SPERT) fuel pins, enriched to 4.81 weight percent ²³⁵U, clad in stainless steel. An experimental apparatus was constructed to hold various

T Trumbull; D Harris

2005-01-01

419

Burning of Minor Actinides and Fission Products from Spent Nuclear Fuel of Power Plants in Dual-Purpose Fusion Reactor  

Microsoft Academic Search

The paper presents the results of analysis of transmutation of Minor Actinides (MA) and Fission Products (FP) from the Spent Nuclear Fuel (SNF) of nuclear power plants. The transmutation scenario includes repeating periods of neutron irradiation in dual-purpose Fusion Power Reactor-Tokamak (FPRT) with Deuterium-Tritium plasma as neutron source and periods of Partitioning and Reprocessing (P&R) of fuel between the irradiation

N. N. Vasiliev; S. V. Sheludjakov; Yu. S. Shpansky; A. G. Serikov

2003-01-01

420

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

SciTech Connect

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

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

2004-03-29

421

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

SciTech Connect

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

Takahashi, H.; Yu, A.

1997-04-01

422

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

Microsoft Academic Search

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (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

J. P. Adams; M. L. Carboneau; D. L. Hagrman

1993-01-01

423

Fission product release from highly irradiated LWR fuel heated to 1300 to 1600°C in steam  

Microsoft Academic Search

Four tests were performed with high-burnup light water reactor (LWR) fuel to explore the amount and characteristics of fission product release at short heating times (0.4 to 10 min) in steam atmosphere in the temperature range 1300 to 1600°C. The test fuel rod segments were cut from full-length fuel rods irradiated at low heat rating to 30,000 MWd\\/MT in the

R. A. Lorenz; J. L. Collins; Malinauskas; M. F. A. P. Osborne; R. L. Towns

1980-01-01

424

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

SciTech Connect

The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 418 nuclides; (2) Covariance uncertainty data for 185 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions at higher energies for isotopes of F, Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new Decay Data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [1].

G. Palmiotti

2011-12-01

425

Water Reactor Safety Research Information Meeting (13th). Volume 6. Fission Product Release and Transport in Containment. Containment Systems Research. Severe Accident Source Term.  

National Technical Information Service (NTIS)

Contents: Fission product release & transport in containment (Status of the ORNL aerosol release and transport project, Status of the DEMONA experiments, Status of the LWR aerosol containment experiments (LACE) program, Experimental validation and improve...

A. J. Weiss

1986-01-01

426

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

SciTech Connect

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

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

1980-04-01

427

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

Code of Federal Regulations, 2013 CFR

...932-235-50-24 (Extractive ActivitiesOil and Gas Topic). Instruction 5 to Item 1204: The average production...computed using production costs disclosed pursuant to FASB ASC Topic 932, Extractive ActivitiesOil and Gas. Units of...

2013-04-01

428

US\\/FRG umbrella agreement for cooperation in GCR development. Fuel, fission products, and graphite subprogram. Quarterly status report, January 1, 1982March 31, 1982  

Microsoft Academic Search

Technical highlights of the period were: Mr. David L. Hanson began a one-year assignment from GA to KFA as a technical specialist working in the area of fission product transport validation; benchmark calculations for verification of fission product release codes were exchanged between GA and HRB (PWS FP-2); fuel capsule R2-K13 was shut down during the entire period while the

1982-01-01

429

Gas utilization platform improves crude oil production  

Microsoft Academic Search

Cooper-Bessemer and CB\\/Southern shipped the 1st of several gas utilization compression platforms late last year to facilitate oil lift and production and conserve natural gas at Venezuela's Lake Maracaibo. The 435-ton package, which was lifted to its platform support, started up, and checked out in 19 days, has been operating since without any shutdown related to equipment problems. The platform

1975-01-01

430

Methane hydrate gas production by thermal stimulation  

SciTech Connect

Two models have been developed to bracket the expected gas production from a methane hydrate reservoir. The frontal-sweep model represents the upper bound on the gas production, and the fracture-flow model represents the lower bound. Parametric studies were made to determine the importance of a number of variables, including porosity, bed thickness, injection temperature, and fracture length. These studies indicate that the hydrate-filled porosity should be at least 15%, reservoir thickness should be about 25 ft or more, and well spacing should be fairly large (maybe 40 acres/well), if possible. Injection temperatures should probably be between 150 and 250/sup 0/F to achieve an acceptable balance between high heat losses and unrealistically high injection rates. Numerous important questions about hydrate gas production remain unanswered.

McGuire, P.L.

1981-01-01

431

Oil and gas leasing/production program  

SciTech Connect

As the Congress declared in the Outer Continental Shelf Lands Act the natural gas and oil production from the Outer Continental Shelf constitutes an important part of the Nation's domestic energy supply. Federal offshore minerals are administered within the Department of the Interior by the Minerals Management Service (MMS), which provides access to potential new sources of natural gas and oil offshore by conducting lease sales. Each year, on or before March 31, the MMS presents to Congress a fiscal year annual report on the Federal offshore natural gas and oil leasing and production program. In FY 1991, this program was the third largest producer of non-tax revenue for the US Treasury, contributing more than $3 billion. This report presents Federal offshore leasing, sales, production, and exploration activities, and environmental monitoring activities.

Heimberger, M.L. (comp.)

1992-03-31

432

Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements  

SciTech Connect

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

Konyashov, Vadim V.; Krasnov, Alexander M. [State Scientific Centre of Russian Federation-Research Institute of Atomic Reactors (Russian Federation)

2002-04-15

433

Decontamination and decommissioning of the fission product pilot plant at the Oak Ridge National Laboratory  

SciTech Connect

The Fission Product Pilot Plant (FPPP) at the Oak Ridge National Laboratory (ORNL) was one of the first facilities used to extract radioactive isotopes from liquid radioactive wastes. During operations, the FPPP was extensively contaminated, resulting in high radiation levels even 30 years after the conclusion of operations. The facility has been abandoned for over 20 years and is now a candidate for decontamination and decommissioning (D&D). In fact, the ORNL management has begun activities toward the D&D of the FPPP. Two of these activities were completed in 1993 and 1994. The first 2030 activity was a facility characterization designed to assess the condition of the interior of the FPPP and to quantify, if possible, the amounts and identities of any radioactive contaminants and hazardous materials as defined by the U.S. Environmental Protection Agency. The facility characterization was intended to determine the condition of the interior, the complement of equipment left in the facility at the time of its closure, and the radiation environment that would be encountered during its D&D. The second activity was an alternatives assessment designed to determine the best approach to the D&D of the FPPP. The alternatives assessment examined five alternatives to decontaminate and decommission the FPPP and recommended the best alternative for its disposition. The first section of the paper describes the FPPP and its history. It includes the various conjectures on what exactly was done when the FPPP was entombed with the shield wall visible in today`s pictures. The next section discusses the characterization that was performed concurrently with the alternatives evaluation. The next two sections detail the D&D plan for the complete dismantelment of the FPPP and its estimated cost and schedule.

Mandry, G.J. [Oak Ridge National Lab., TN (United States); Snedaker, W. [Enserch Environmental Corp., Oak Ridge, TN (United States)

1994-11-01

434

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

435

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

436

Conceptual Analysis of the Power Production of Fission Electric Cell Reactors  

SciTech Connect

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

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

2002-07-01

437

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

438

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

Microsoft Academic Search

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

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

2007-01-01

439

Fission fusion hybrids- recent progress  

NASA Astrophysics Data System (ADS)

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

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

2012-03-01

440

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

Microsoft Academic Search

Economic studies on simulated gas hydrate reservoirs have been compiled to estimate the price of natural gas that may lead to economically viable production from the most promising gas hydrate accumulations. As a first estimate, $CDN2005 12\\/Mscf is the lowest gas price that would allow economically viable production from gas hydrates in the absence of associated free gas, while an

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

2009-01-01

441

Oil and Gas Leasing\\/Production Program  

Microsoft Academic Search

As the Congress declared in the Outer Continental Shelf Lands Act (OCSLA), the natural gas and oil production from the Outer Continental Shelf (OCS) constitutes an important part of the Nation's domestic energy supply. Federal offshore minerals are administered within the Department of the Interior by the Minerals Management Service (MMS), which provides access to potential new sources of natural

M. L. Heimberger; D. OBrien

1991-01-01

442

New Methodology for Natural Gas Production Estimates  

EIA Publications

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

Information Center

2010-04-26

443

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; . ro

2003-01-01

444

Separation processes for rare gas fission products from the off-gas of nuclear facilities  

Microsoft Academic Search

With the advance of atomic energy utilization, the problem of radiation ; exposure and its effect is attracting world-wide interest. It is necessary to ; control release of radioactive gases from nuclear power plants and fuel ; reprocessing plants into the atmosphere. Of rare gases from the nuclear power ; plants, ⁸⁵Kr is the most important because of its long

1973-01-01

445

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

SciTech Connect

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

Gohar, Y.

2000-11-15

446

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

2012-09-03

447

Postirradiation dimensional stability and fission product behavior of deliberately defected UO2 fuel at 200 and 400C  

SciTech Connect

Twenty-four-hour sweep tests have been carried out in flowing air at 200 and 400C on deliberately defected UO2 fuel elements with 2.5-yr discharge times. At 200C there was no diametral change, but at 400C, swelling and severe sheath cracking were observed. Neither short-lived fission products nor TUCs, TXCs, or WRu were detected above background. Maximum YVKr release was less than or equal to 7.4 X 10U Bq (less than or equal to 2 X 10 W Ci).

Hustings, I.J.; Barrand, R.D.; Klem, J.R.; McCracken, D.R.; Mizzan, E.; Nash, K.E.; Novak, J.

1985-08-01

448

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

449

Book Review: Fission product transport processes in reactor accidents, proceedings to the International Centre for Heat and Mass processes  

SciTech Connect

In the wake of the Chernobyl accident, the Executive Committee of the International Centre for Heat and Mass Transfer decided at its meeting in Tashkent in April 1987 to organize a seminar on fission product transport processes in reactor accidents. The seminar was held in Dubrovnik, Yugoslavia, May 22-26, 1989. Co-sponsoring organizations were the United Nations Educational Scientific and Cultural Organization, Carleton University, and the Boris Kidric Institute of Nuclear Sciences. Twelve invited lectures and forty-seven invited papers were presented. 1 tab.

Kress, T.S. [Martin Marietta Energy Systems, Oak Ridge, TN (United States)

1991-01-01

450

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

451

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

Microsoft Academic Search

A fuel rod has been instrumented with a new design of an acoustic resonator used to measure in a non destructive way the internal rod plenum gas mixture composition. This ultrasonic sensor has demonstrated its ability to operate in pile during REMORA 3 irradiation experiment carried out in the OSIRIS Material Testing Reactor (CEA Saclay, France). Due to very severe

E. Rosenkrantz; J. Y. Ferrandis; F. Augereau; T. Lambert; D. Fourmentel; X. Tiratay

2011-01-01

452

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

SciTech Connect

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

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

2010-09-23

453

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

454

Yrast Excitations around Doubly Magic 132Sn from Fission Product gamma-Ray Studies  

Microsoft Academic Search

Prompt gamma-ray cascades in neutron-rich nuclei around doubly magic 132Sn have been studied at Eurogam II using a 248Cm fission source. Yrast states to above 5.5 MeV in the two- and three-proton N = 82 isotones 134Te and 135I are reported. They are interpreted in terms of valence proton and particle-hole core excitations with the help of shell model calculations

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

1996-01-01

455

Recovery of fission product rare earth sulfates from Purex 1WW  

Microsoft Academic Search

Cerium-144 and promethium-147, accompanied by rare earths resulting from fission or decay can be removed from Purex 1WW in >90% yield as an insoluble, crystalline sodium-rare earth double sulfate. Precipitation is initiated by a one-to-three hour equilibration at 90°C and centrifugation at 90°C to take advantage of the lower solubility of the double sulfate salt at a higher temperature. The

E. J. Wheelwright; W. H. Swift

1961-01-01

456

THE RECOVERY OF FISSION PRODUCT RARE EARTH SULFATES FROM PUREX 1WW  

Microsoft Academic Search

Cerium- and 144 promethium-147, accompanied by rare earths resulting ;\\u000a from fission or decay can be removed from Purex 1WW in>90% yield as an ;\\u000a insoluble, crystalline sodium-rare earth double sulfate. Precipitation is ;\\u000a initiated by a one-to-three hour equilibration at 90 deg C and centrifugation at ;\\u000a 90 deg C to take advantage of the lower solubility of the

E. J. Wheelwright; W. H. Swift

1961-01-01

457

Natural gas production, proration and markets  

SciTech Connect

I would like to give you just a few of the technical aspects of what the Commission has done in its recent amendment of the gas proration rules. First, the Commission replaced the historical pipeline estimations of market demand with a process that includes what we call an optional producer forecast. This begins with the Commission creating a market demand forecast for prorated gas fields based on last year's production and mailing it to each operator, showing the amount of gas that was produced from that operator's wells in each field. The operator can then review that forecast. If it is a satisfactory estimate of market demand, then the operator need not respond. however, if indeed the market demand has gone up or down from last year's level, then the operator has the option of filling out an optional market demand forecast form that will override the Commission's estimate. That is the first significant change in the new rules.

Garlick, D.M. (Railroad Commission of Texas, Austin (United States))

1992-06-01

458

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

459

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

460

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

461

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

SciTech Connect

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

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

2011-09-23

462

Gas production in the MEGAPIE spallation target  

SciTech Connect

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

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

2011-07-01

463

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

SciTech Connect

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

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

2011-12-01

464

Radiolytic gas production from tritiated waste forms  

Microsoft Academic Search

Radiolytic gas production during long-term storage of tritiated waste was estimated from gamma and alpha radiolysis tests to determine the extent of pressurization in sealed containers. Two forms of simulated wastes were irradiated with ⁶°Co gamma rays or ²⁴⁴Cm alpha particles: concrete for solidification of tritiated water and vermiculite for solidification of tritiated octane or vacuum pump oil. For concrete,

N. E. Bibler; E. G. Orebaugh

1977-01-01