22 Matching Results

Search Results

Advanced search parameters have been applied.

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

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

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

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

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

Calculation of the evolution of the fuel microstructure in UMo alloys and implications for fuel swelling.

Description: The evolution of a cellular dislocation structure and subsequent recrystallization have been identified as important aspects of the irradiated UMo alloy microstructure that can have a strong impact on dispersion fuel swelling. Dislocation kinetics depends on the preferential bias of dislocations for interstitial compared to vacancies. This paper presents theoretical calculations for the evolution of a cellular dislocation structure, and recrystallization in U-10Mo. Implications for fuel swelling are discussed.
Date: October 1, 1999
Creator: Rest, J.; Hofman, G. L.; Konovalov, I. & Maslov, A.
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

DART model for thermal conductivity of U{sub 3}Si{sub 2} aluminum dispersion fuel

Description: This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminium dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values.
Date: September 1, 1995
Creator: Rest, J.; Snelgrove, J.L. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Development and Verification of the LIFE-GCFR Computer Code for Predicting Gas-Cooled Fast-Reactor Fuel-Rod Performance

Description: The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects.
Date: December 1980
Creator: Hsieh, T. C.; Billone, Michael C. & 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

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 1995
Creator: Rest, J.
Partner: UNT Libraries Government Documents Department

Aluminum-U{sub 3}Si{sub 2} interdiffusion and its implications for the performance of highly loaded fuel operating at higher temperatures and fission rates

Description: Recent irradiation tests of U{sub 3}Si-Al dispersion fuel have shown performance limitations of this fuel when high volume fractions of U{sub 3}Si{sub 2} operate at high temperatures and high fission rates. This potential problem is associated with high rates of Al-U{sub 3}Si{sub 2} interdiffusion that may lead to complete consumption of matrix aluminum and the formation of excessive porosity.
Date: December 1, 1996
Creator: Hofman, G.L.; Rest, J.; Snelgrove, J.L.; Wiencek, T. & Koster van Groos, S.
Partner: UNT Libraries Government Documents Department

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

Description: The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.
Date: October 18, 2007
Creator: Kaganas, G. & Rest, J.
Partner: UNT Libraries Government Documents Department

P-West High Intensity Secondary Beam Area Design Report

Description: This report gives the initial design parameters of a 1000 GeV High Intensity Superconducting Secondary Beam Laboratory to be situated in the Proton Area downstream of the existing Proton West experimental station. The area will provide Fermilab with a major capability for experimentation with pion and antiproton beams of intensities and of energies available at no other laboratory and with an electron beam with excellent spot size, intensity, and purity at energies far above that available at electron machines. Detailed beam design, area layouts, and cost estimates are presented, along with the design considerations.
Date: March 1, 1977
Creator: Cox, A.; Currier, R.; Eartly, D.; Guthke, A.; Johnson, G.; Lee, D. et al.
Partner: UNT Libraries Government Documents Department

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

Description: Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up to ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.
Date: November 1, 2009
Creator: Gan, J.; Keiser, D.; Wachs, D.; Miller, B.; Allen, T.; Kirk, M. et al.
Partner: UNT Libraries Government Documents Department