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Modeling studies on the precipitation of Kr after implantation into metals

Description: A rate-theory approach is applied to interpreting observations on the precipitation of Kr injected into Ni at temperatures between 25 and 560/degree/C. At temperatures of 400/degree/C or higher, the implanted Kr precipitates evolve into a bi-modal size distribution containing small solid precipitates and an additional population of larger, faceted bubbles. The calculations explore the dependence of the observed bi-modal distribution on the maximum size of the solid Kr precipitates and the effect of this dependence on bubble mobility. The analysis suggests that during the irradiation, whereas the large bubbles move by surface diffusion, the solid Kr precipitates are immobile. The relevance of the Kr-Ni interaction on the solid Kr precipitates size cutoff is discussed. 18 refs., 8 figs., 2 tabs.
Date: February 1, 1988
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

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

Description: The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted.
Date: October 1, 1985
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Kinetics of fission product release prior to fuel slumping

Description: This paper describes the primary physical/chemical models recently incorporated into a mechanistic code (FASTGRASS) for the estimation of fission product release from fuel, and compares predicted results with test data. The theory of noble gas behavior is discussed in relation to its effect on the release behavior of I, Cs, Te, Ba, and Sr. The behavior of these fission products in the presence of fuel liquefaction/dissolution and oxidation grain-growth phenomena is presented, as is the chemistry of Sr, Ba, I, and Cs. Comparison of code predictions with data indicates the following trends. Fission product release behavior from solid strongly depends on fuel microstructure, irradiation history, time at temperature, and internal fuel rod chemistry. Fuel liquefaction/dissolution, fracturing, and oxidation also exert a pronounced effect on release during fuel rod degradation. For very low burnup fuel appreciable fission product retention in previously liquefied fuel can occur due to the low concentration of fission products, and the limited growth of bubbles in the liquefied material. 24 refs., 13 figs., 9 tabs.
Date: October 1, 1987
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Prediction of the response of fission-gas in nuclear fuels during normal and transient heating conditions

Description: A discussion is presented on the interplay between mechanisms of fission-gas behavior, and on how the assumed relationships between various mechanisms affect predictions for the behavior of the gas during a wide range of operating conditions. Highlighted in the discussion are the relationships between intragranular fission gas mobilities during normal and off-normal (nonequilibrium) conditions and intra- and intergranular gas-atom re-solution. The main conclusion reached is that more definitive information about the nature of the mechanisms and about the synergistic relationship between them await further experiments and analyses of the observed phenomena.
Date: January 1, 1979
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

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

Description: A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.
Date: September 1, 1983
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

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

Description: A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles.
Date: September 1, 1983
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Scoping analysis of fission gas behavior in UO/sub 2/ fuel for an in-core thermionic reactor

Description: This paper gives results from a preliminary evaluation of swelling, and, in particular, swelling caused by the retained fission gases (Xe, Kr) in the fuel. Design parameters of the thermionic unit cell and its operating conditions were provided by Space Power, Inc. The analysis tool used is a mechanistic fission-gas behavior code, FASTGRASS, developed by J. Rest at ANL (Rest, 1984). In the FASTGRASS analysis, the UO/sub 2/ fuel column is taken as a solid cylinder. No external restraint is imposed on the UO/sub 2/ fuel other than the ambient pressure. The maximum fuel burnup is assumed to be 5 at%. The operating conditions specified for the UO/sub 2/ fuel fall into two categories: (a) low linear power (q' = 5-8 kW/ft) and high fuel surface temperature (T/sub s/ = 1700, 1800, 1900 K) and (b) high linear power (q' = 8-12 kW/ft) and low fuel surface temperature (T/sub s/ = 1300, 1400, 1500, 1600 K). Taking the upper and lower linear powers in each category in combination with the specified fuel surface temperatures resulted in a total of 14 cases for the scoping analysis. The total irradiation time (based on linear power and 5 at% burnup) for these cases varies from 2.5 to 6 years.
Date: October 1, 1984
Creator: Liu, Y.Y. & Rest, J.
Partner: UNT Libraries Government Documents Department

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

Description: GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included in both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.
Date: January 1, 1979
Creator: Rest, J. & Gehl, S.M.
Partner: UNT Libraries Government Documents Department

Mechanistic model for Sr and Ba release from severely damaged fuel

Description: Among radionuclides associated with fission product release during severe accidents, the primary ones with health consequences are the volatile species of I, Te, and Cs, and the next most important are Sr, Ba, and Ru. Considerable progress has been made in the mechanistic understanding of I, Cs, Te, and noble gas release; however, no capability presently exists for estimating the release of Sr, Ba, and Ru. This paper presents a description of the primary physical/chemical models recently incorporated into the FASTGRASS-VFP (volatile fission product) code for the estimation of Sr and Ba release. FASTGRASS-VFP release predictions are compared with two data sets: (1) data from out-of-reactor induction-heating experiments on declad low-burnup (1000 and 4000 MWd/t) pellets, and (2) data from the more recent in-reactor PBF Severe Fuel Damage Tests, in which one-meter-long, trace-irradiated (89 MWd/t) and normally irradiated (approx.35,000 MWd/t) fuel rods were tested under accident conditions. 10 refs.
Date: November 1, 1985
Creator: Rest, J. & Cronenberg, A.W.
Partner: UNT Libraries Government Documents Department

Application of a mechanistic model for radiation-induced amorphization and crystallization of uranium silicide to recrystallization of UO{sub 2}

Description: An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO{sub 2} wherein the stored energy in the material is concentrated in a network of sinklike nuclei that diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. A generalized theory of radiation-induced amorphization and crystallization, developed for intermetallic nuclear materials, is applied to UO{sub 2}. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission- induced damage to the material.
Date: July 1, 1996
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

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

Description: This report describes the primary physical models that form the basis of the DART mechanistic computer model for calculating fission-product-induced swelling of aluminum dispersion fuels; the calculated results are compared with test data. In addition, DART calculates irradiation-induced changes in the thermal conductivity of the dispersion fuel, as well as fuel restructuring due to aluminum fuel reaction, amorphization, and recrystallization. Input instructions for execution on mainframe, workstation, and personal computers are provided, as is a description of DART output. The theory of fission gas behavior and its effect on fuel swelling is discussed. The behavior of these fission products in both crystalline and amorphous fuel and in the presence of irradiation-induced recrystallization and crystalline-to-amorphous-phase change phenomena is presented, as are models for these irradiation-induced processes.
Date: August 1, 1995
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Experimental and calculated swelling behavior of U-10 wt.% Mo under low irradiation temperatures.

Description: SEM micrographs of U-10 wt.% Mo irradiated at low temperature in the ATR to about 40 at. % burnup show the presence of cavities. We have used a rate-theory-based model to investigate the nucleation and growth of cavities during low-temperature irradiation of uranium-molybdenum alloys in the presence of irradiation-induced interstitial-loop formation and growth. In addition, the evolution of forest dislocations was calculated based on dislocation loop growth and simultaneous climb and glide of unfaded loops. Consolidation of the dislocation structure takes into account capture of interstitial dislocation loops and annihilation of adjacent dislocations, as well as loss to grain boundaries. A di-interstitial is assumed to be the nucleus of a dislocation loop. Cavities are nucleated when two gas atoms come together in the presence of at least one vacancy. Cavity growth occurs by the influx of gas atoms and/or vacancies. In turn, the free interstitial concentration, and thus (due to recombination) the free-vacancy concentration, depends on the dislocation density. Bias-driven growth of cavities can lead to substantial swelling of the alloy (void swelling). However, our calculations indicate that the swelling mechanism in the U-10 wt.% Mo alloy at low irradiation temperatures is fission gas driven. The calculations also indicate that the observed bubbles must be associated with a sub-grain structure. Calculated swelling and bubble-size-distribution are compared with irradiation data.
Date: September 29, 1998
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Fission gas bubble nucleated cavitational swelling of the alpha-uranium phase of irradiated U-Pu-Zr fuel

Description: Cavitational swelling has been identified as a potential swelling mechanism for the alpha uranium phase of irradiated U-Pu-Zr metal fuels for the Integral Fast Reactor being developed at Argonne National Laboratory. The trends of U-Pu-Zr swelling data prior to fuel cladding contact can be interpreted in terms of unrestrained cavitational driven swelling. It is theorized that the swelling mechanisms at work in the alpha uranium phase can be modeled by single vacancy and single interstitial kinetics with intergranular gas bubbles providing the void nuclei, avoiding the use of complicated defect interaction terms required for the calculation of void nucleation. The focus of the kinetics of fission gas evolution as it relates to cavitational swelling is prior to the formation of a significant amount of interconnected porosity and is on the development of small intergranular gas bubbles which can act as void nuclei. Calculations for the evolution of intergranular fission gas bubbles show that they provide critical cavity sizes (i.e., the size above which the cavity will grow by bias-driven vacancy flux) consistent with the observed incubation dose for the onset of rapid swelling and gas release.
Date: April 1, 1992
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

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

Description: The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures.
Date: October 1, 1983
Creator: Rest, J.; Zawadski, S.A. & Piasecka, M.
Partner: UNT Libraries Government Documents Department

FastDart : a fast, accurate and friendly version of DART code.

Description: A new enhanced, visual version of DART code is presented. DART is a mechanistic model based code, developed for the performance calculation and assessment of aluminum dispersion fuel. Major issues of this new version are the development of a new, time saving calculation routine, able to be run on PC, a friendly visual input interface and a plotting facility. This version, available for silicide and U-Mo fuels,adds to the classical accuracy of DART models for fuel performance prediction, a faster execution and visual interfaces. It is part of a collaboration agreement between ANL and CNEA in the area of Low Enriched Uranium Advanced Fuels, held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy.
Date: November 8, 2000
Creator: Rest, J. & Taboada, H.
Partner: UNT Libraries Government Documents Department

Dart model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

Description: The Dispersion Analysis Research Tool (DART) contains models for fission-gas induced fuel swelling, interaction of fuel with the matrix aluminum, resultant reaction-product swelling, and calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U{sub 3}SiAl-Al and U{sub 3}Si{sub 2}-Al for various dispersion fuel element designs with the data. DART results are compared with data for fuel swelling Of U{sub 3}SiAl-Al in plate, tube, and rod configurations as a function of fission density. Plate and tube calculations were performed at a constant fuel temperature of 373 K and 518 K, respectively. An irradiation temperature of 518 K results in a calculated aluminide layer thickness for the Russian tube that is in the center of the measured range (16 {mu}m). Rod calculations were performed with a temperature gradient across the rod characterized by surface and central temperatures of 373 K and 423 K, respectively. The effective yield stress of irradiated Al matrix material and the aluminide was determined by comparing the results of DART calculations with postirradiation immersion volume measurement of U{sub 3}SiAl plates. The values for the effective yield stress were used in all subsequent simulations. The lower calculated fuel swelling in the rod-type element is due to an assumed biaxial stress state. Fuel swelling in plates results in plate thickness increase only. Likewise, in tubes, only the wall thickness increases. Irradiation experiments have shown that plate-type dispersion fuel elements can develop blisters or pillows at high U-235 burnup when fuel compounds exhibiting breakaway swelling are used at moderate to high fuel volume fractions. DART-calculated interaction layer thickness and fuel swelling follows the trends of the observations. 3 refs., 2 figs.
Date: June 1, 1997
Creator: Rest, J. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Behavior of irradiated LWR fuel pellets during thermal transients

Description: Prediction of the behavior of LWR fuel rods and fission products under off-normal and accident conditions requires a physically realistic description of fuel swelling and fission-product release that currently does not exist. To satisfy this need, a program was initiated at ANL approximately a year ago with the prime objective of developing a comprehensive computer-base model that describes the release of fission products as a function of thermal transients anticipated in hypothetical accident situations. This model will be incorporated into ANC's FRAP accident-analysis code system. The analytical effort is supported by data developed from characterization of irradiated LWR fuel and from out-of-reactor transient heating tests of irradiated LWR fuel under conditions that simulate hypothetical LWR accidents. (auth)
Date: January 1, 1975
Creator: Kelman, L.R.; Rest, J.; Seitz, M.G. & Gehl, S.M.
Partner: UNT Libraries Government Documents Department

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

Description: 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 ...
Date: April 14, 2008
Creator: Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.; Division, Nuclear Engineering & RIAR, SSCR
Partner: UNT Libraries Government Documents Department

Effect of recrystallization in high-burnup UO{sub 2} on gas release during RIA-type transients

Description: The authors recently proposed a model for irradiation-induced recrystallization (grain subdivision) and swelling in UO{sub 2} fuels wherein the stored energy in the material is concentrated in a network of sink-like nuclei that diminish with dose due to interaction with radiation-produced defects. It is of considerable interest to explore the consequences of recrystallization on gas release during a reactivity initiated accident (RIA). In the absence of recrystallization, gas release during RIA-type transients is generally limited to gas available on grain boundaries and edges due to the very short heatup times (milliseconds), short cooldown periods (seconds), and relatively long intragranular diffusion distances (on the order of micrometers). However, recrystallization provides grain-boundary surfaces that are substantially closer to the gas retained in the bulk material, and thus the potential for much higher gas release. The authors show the calculated burnup at which grain subdivision will occur as a function of fractional radius and fuel temperature for a generic pressurized water reactor irradiation. The FASTGRASS code was used to calculate fission gas behavior during in-reactor irradiation and during the RIA-type transient. Results are given. It is clear from these results that recrystallization of high-burnup UO{sub 2} has implications for the potential consequences of severe accident scenarios such as the RIA type.
Date: October 1, 1994
Creator: Rest, J. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Irradiation behavior of uranium oxide-aluminum dispersion fuel

Description: An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO{sub 2}-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U{sub 3}O{sub 8}-Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U{sub 3}O{sub 8} are valid for UO{sub 2}, the LEU UO{sub 2}-Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10{sup 27} fissions m{sup {minus}3} ({approximately} 63% {sup 235}U burnup).
Date: December 1, 1996
Creator: Hofman, G.L.; Rest, J. & Snelgrove, J.L.
Partner: UNT Libraries Government Documents Department

Analysis of the swelling behavior of U-alloys

Description: Available data on two alloys from the EBR-II driver fuel development program have been utilized in the construction and validation of mechanistic models aimed at elucidating swelling mechanisms in high density uranium alloys. Swelling predictions are made under ATR conditions for U-10Mo fuels, currently under irradiation in the ATR, and for U-10Zr.
Date: October 1, 1997
Creator: Rest, J.; Hofman, G.L.; Coffey, K.L.; Konovalov, I. & Maslov, A.
Partner: UNT Libraries Government Documents Department